Method of processing nitrate-containing radioactive liquid wastes

International Nuclear Information System (INIS)

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

1983-01-01

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

Management of wastes containing radioactivity from mining and milling of uranium ores in Northern Australia

International Nuclear Information System (INIS)

Costello, J.M.

1977-01-01

The procedures and controls to achieve safe management of wastes containing radioactivity during the mining and processing of uranium ores are mainly site specific depending on the nature, location and distribution of the ore and gangue material. Waste rock and below-ore-grade material containing low levels of radioactivity require disposal at the mine site. In open cut mining the material is generally stockpiled above ground, with revegetation and collection of run-off water. Some material may be used to backfill open cuts. Management of these wastes requires a thorough investigation of ground water hydrology and surface soil characteristics to control dissipation of radioactive material. Dust containing radon and radioactive particulate is produced during ore milling, and dusts of ore concentrate are generated during calcination and packaging of the yellowcake product. These dusts are managed by ventilation and filtration systems, working conditions, and discharges to atmosphere will be according to the Australian Code of Practice on Radiation Protection during Mining and Milling of Uranium Ores. The chemical waste stream from leaching and processing of the uranium ores contains the majority of the radioactivity resulting from radium and its decay products. Neutralised effluent is discharged into holding ponds for settling of solids. This paper describes the nature of wastes containing radioactivity resulting from the mining and milling of uranium, and illustrates modern engineering practices and monitoring procedures to manage the wastes, as described in the Environmental Impact statement produced by Ranger Uranium Mines Proprietary Limited for public hearings

Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

International Nuclear Information System (INIS)

Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

2000-01-01

A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

Process for treatment of detergent-containing radioactive liquid wastes

International Nuclear Information System (INIS)

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

1984-01-01

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

The Separation Of The Uranium On The Liquid Radioactive Wastes By Using Zeolites

International Nuclear Information System (INIS)

Budiyono, ME; Peornomo, Herry; S, Djoko

2000-01-01

The experimental investigation on the separation of uranium of liquid radioactive wastes by using zeolite has been carried out. The aims of this investigation was to separate uranium and to determine the solid content of the liquid radioactive waste. This investigation was carried out because zeolite could be used as a good sorption material. The investigation parameters were the number of zeolites, the pH of wastes and the grain size of zeolites. The number of zeolites was varied from 10-100 g, the waste pH was varied from 1-12 and the grain size of zeolites was varied from 20/40-100/200 mesh, each parameters to included in the waste. The conclusion that could be drawn from this investigation were that the best result of the 80 g zeolites used, pH was 9, and the grain size of zeolites was 60/80 mesh, the solid content was 119,46 g/l and efficiency of separation was 81,74 %. As a rule, the solid content in the evaporator process should be about 200-300 g/l

Transport of uranium by supported liquid membrane containing bis(2-ethylhexyl) hydrogenphosphate and 1-octanol

International Nuclear Information System (INIS)

Akiba, Kenichi; Kanno, Takuji; Takahashi, Toshihiko.

1984-01-01

Carrier-mediated transport of uranium(VI) has been studied by means of liquid membranes impregnated in a microporous polymer. Liquid membranes containing bis(2-ethylhexyl) hydrogenphosphate (DEHPA) alone yielded inadequate stripping of uranium. The addition of 1-octanol to DEHPA solutions resulted in a decrease in extractability, and made it possible to control the distribution ratio of uranium. Uranium in the feed solution was sufficiently transported across the liquid membrane containing this DEHPA-1-octanol mixture into the product solution. The apparent rate constant (ksub(obs)) of transport increased slightly with an increase in carrier concentrations. Variations in acid concentrations of the feed solution (pH 2.5--3.2) and the product solution (0.1--1.0 M H 2 SO 4 ) had little effect on the transport rate. A large excess of uranium, more than the carrier content in the liquid membrane, was finally concentrated in the stripping acid. (author)

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

International Nuclear Information System (INIS)

Wakaida, Yasuo.

1996-01-01

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

Development of a process to reduce the uranium concentration of liquid radioactive waste

International Nuclear Information System (INIS)

Fuentealba Toro, Edgardo David

2015-01-01

The purpose of radioactive waste management is to prevent the discharge of waste into the biosphere, a function carried out in Chile by the Chilean Nuclear Energy Commission (CCHEN), which stores around 500 [L] of these organic and inorganic waste in cans coming from research of Universities and CCHEN' laboratories. Within the inorganic liquid waste are concentrations of Uranyl salts with sulfates, chlorides and phosphates. The purpose of this work is to develop at laboratory level a process to concentrate and precipitate uranium salts (Sulfate and Uranyl Chloride) present in radioactive liquid effluents, because in the case of these very long period wastes in liquid state, the most widely used processes are aimed at concentrating or extracting radioactive compounds through separation processes, for their conditioning and final storage under conditions whose radiological risk is minimized. The selected process is liquid-liquid extraction, being evaluated solvents such as benzene and kerosene with the following extractants: tri-n-octylphosphine oxide (TOPO), di-2-ethylhexyl phosphoric acid (DEHPA) and Cyanex© 923. To determine the extraction conditions, which allow to reduce the concentration of uranium to values lower than 10 ppm, the extractant concentration was modified from 0.05 to 0.41 [M] with solvent volume / residue (VO/VA) ratios of 0.2 to 0.5, at an initial concentration of 8,446 [gU/L] and subsequent precipitation of uranium extracted by a reaction with ammonium carbonate. From these experimental tests the maximum extraction conditions were determined. To the generated effluents, a second stage of extraction was necessary in order to reduce its concentration below 10 [mg / L]. The experimental tests allowed to reduce the concentration under 2.5 [mgU/L], equivalent to 99.97% extraction efficiency. The tests with Cyanex© 923 in replacement of the TOPO, allowed to obtain similar results and even better in some cases, due to the fact that final

Plant start-up operation of treatment of liquid radioactive waste containing uranium in sulfuric acid by means of ion exchange

International Nuclear Information System (INIS)

Bustamante Escobedo, Mauricio

2003-01-01

The hydrometallurgy of uranium developed by CCHEN in the '80s generated different types of liquid radioactive waste: organic and aqueous, which have to be treated to extract the uranium, obtaining uranium concentrate less than 3 [ppm], in order to comply with the current regulations and to be able to be discharged as industrial waste over to specialized companies. The radioactive wastes were treated using strongly basic ion exchange resins to obtain a lower volume of concentrated uranium in solid phase. Batch processes were carried out with different pH values (1.4-1.6-1.7-1-8 and 2.0) and different residence times (5-10-15-240 [min]), to define the optimum processing conditions. Three resins were used: IONAC, AFP-329, IONAC A-641, DOWEX 11. The load capacities for each one were: 2[g/l], 1.85 [g/l] and 2.1 [g/l], respectively. Dowex 11 obtained the best uranium load results. Because of the nature of the treated solutions, the pH variation with sodium hydroxide generated precipitation and then crystallization of the existing pieces. Ninety-five percent of the resin's maximum load was attained after about 20 [min] of residence time. Continuous tests were carried out at the solution's original pH to verify the results obtained in the batch stage. This ion exchange process was compared with the chemical precipitation, with lime and sodium hydroxide as well as evaporation and in both processes the volumes of solid that were generated were significantly greater. The ion exchange plant worked with two serially connected columns, using 10 [l] of Dowex 11 resin in each one, with a residence time of 20 [min] per column. A total of 8 [m 3 ] of solution, obtaining an effluent with a uranium concentrate of less than 3 [ppm], using 140 [l] of resin, which generated an average load of 3 [g/l] of uranium (au)

Application of insoluble tannin to recovery of uranium, TRU and heavy metals elements form radioactive liquid waste

International Nuclear Information System (INIS)

Hamaguchi, Kazuhiko; Shirato, Wataru; Nakamura, Yasuo; Matsumura, Tatsuro; Takeshita, Kenji; Nakano, Yoshio

1999-01-01

Mitsubishi Nuclear Fuel Co., Ltd. (MNF) has developed a new adsorbent, TANNIX (tread mark), for the recovery of uranium, TRU and heavy metal elements in the liquid waste, in which TANNIX derived from a natural tannin polymer. TANNIX has same advantages that handling is easier than that of standard IX-resin, and that the volume of secondary waste is reduced by burning the used TANNIX. We have replaced its radioactive liquid waste treatment system from the conventional co-precipitation process to adsorption process by using TANNIX. TANNIX was founded to be more effective for the recovery of Pu, TRU, and hexavalent chromium Cr-(VI) as well as Uranium. (author)

Separation of uranium and common impurities from solid analytical waste containing plutonium

International Nuclear Information System (INIS)

Pathak, Nimai; Kumar, Mithlesh; Thulasidas, S.K.; Hon, N.S.; Kulkarni, M.J.; Mhatre, Amol; Natarajan, V.

2014-07-01

The report describes separation of uranium (U) and common impurities from solid analytical waste containing plutonium (Pu). This will be useful in recovery of Pu from nuclear waste. This is an important activity of any nuclear program in view of the strategic importance of Pu. In Radiochemistry Division, the trace metal analysis of Pu bearing fuel materials such as PuO 2 , (U,Pu)O 2 and (U,Pu)C are being carried out using the DC arc-Carrier Distillation technique. During these analyses, solid analytical waste containing Pu and 241 Am is generated. This comprises of left-over of samples and prepared charges. The main constituents of this waste are uranium oxide, plutonium oxide and silver chloride used as carrier. This report describes the entire work carried out to separate gram quantities of Pu from large amounts of U and mg quantities of 241 Am and the effect of leaching of the waste with nitric acid as a function of batch size. The effect of leaching the solid analytical waste of (U,Pu)O 2 and AgCl with concentrated nitric acid for different time intervals was also studied. Later keeping the time constant, the effect of nitric acid molarity on the leaching of U and Pu was investigated. Four different lots of the waste having different amounts were subjected to multiple leaching with 8 M nitric acid, each for 15 minutes duration. In all the experiments the amount of Uranium, Plutonium and other impurities leached were determined using ICP as an excitation source. The results are discussed in this report. (author)

Trace recovery of uranium and rare earth contained in phosphates by liquid-liquid extraction in sulfuric attack liquor

International Nuclear Information System (INIS)

Bousquet, F.; Foraison, D.; Leveque, A.; Sabot, J.L.

1980-06-01

Uranium and rare earths can be recovered in sedimentary phosphates during the wet processing of the ore by sulfuric acid giving raw phosphoric acid at 30 per cent of P 2 O 5 . Practically all the uranium contained and only part of rare earths are put into solution in this treatment. Separation of these elements in the phosphoric solution is obtained by liquid-liquid extraction with alkylphosphoric acids and especially with their mono and di esters. Partition isotherms are determined and counter-current tests are effected. Uranium and rare earths reextraction from these solvents can be simultaneous or separate with aqueous solutions alkaline or containing HF or by antisynergism. Pros and cons of each reextraction process are discussed. In conclusion HDEHP or OPPA are recommended because of availability, stability and hydrodynamic, OPPA less selective with rare earths allows the recovery with uranium of ceric earths, yttrium and yttric earths [fr

Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

International Nuclear Information System (INIS)

Ghaib Widodo; Bambang Herutomo

2007-01-01

Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

Method of processing liquid waste containing fission product

International Nuclear Information System (INIS)

Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

1988-01-01

Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

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.)

Recovery of uranium from analytical waste solution

International Nuclear Information System (INIS)

Kumar, Pradeep; Anitha, M.; Singh, D.K.

2016-01-01

Dispersion fuels are considered as advance fuel for the nuclear reactor. Liquid waste containing significant quantity of uranium gets generated during chemical characterization of dispersion fuel. The present paper highlights the effort in devising a counter current solvent extraction process based on the synergistic mixture of D2EHPA and Cyanex 923 to recover uranium from such waste solutions. A typical analytical waste solution was found to have the following composition: U 3 O 8 (∼3 g/L), Al: 0.3 g/L, V: 15 ppm, Phosphoric acid: 3M, sulphuric acid : 1M and nitric acid : 1M. The aqueous solution is composed of mixture of either 3M phosphoric acid and 1M sulphuric acid or 1M sulphuric acid and 1M nitric acid, keeping metallic concentrations in the above mentioned range. Different organic solvents were tested. Based on the higher extraction of uranium with synergistic mixture of 0.5M D2EHPA + 0.125M Cyanex 923, it was selected for further investigation in the present work

Removal of Radioactive Pollutants by Liquid Emulsion Membrane From Liquid Waste

International Nuclear Information System (INIS)

Yossef, Y.A.A.

2013-01-01

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

Lime treatment of liquid waste containing heavy metals, radionuclides and organics

International Nuclear Information System (INIS)

DuPont, A.

1990-01-01

This paper reports on lime treatment of liquid waste containing heavy metals, radio nuclides and organics. Lime is wellknown for its use in softening drinking water the treatment of municipal wastewaters. It is becoming important in the treatment of industrial wastewater and liquid inorganic hazardous waste; however, there are many questions regarding the use of lime for the treatment of liquid hazardous waste

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, Wilbur O.

1985-01-01

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

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, W.O.

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

Waste monitoring of the uranium ore processing activities in Romania

International Nuclear Information System (INIS)

Nica, L.

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Laboratory studies on leaching of low grade uranium ores and treatment of low level liquid waste generated by leaching experiments

International Nuclear Information System (INIS)

Palabrica, O.T.; Antonino, E.J.; Caluag, L.A.; Villamater, D.

1980-07-01

Acid leaching experiments of preconcentrated uranium ore were carried out at a pulp density of 50% solids, using sulfuric acid with sodium chlorate as oxidant. The different leaching parameters considered in this work were temperature, oxidant level and leaching time. In the experimental procedure, the concentration of oxidant and the temperature were varied to determine how they affect the leaching process. Experimental results are illustrated in tabulated form for better interpretation. Uranium analyses were done by fluorimetric and delayed-neutron activation analysis. An anion exchange method using Dowex 1 x 8, 200-400 mesh (Cl - ) was used in treating the low-level liquid waste generated by leaching experiments. The purpose of this treatment was to minimize radioactive contamination in the waste materials and also to recover some of the uranium left in the liquid waste. (author)

Recovery and removal of uranium by using plant wastes

International Nuclear Information System (INIS)

Nakajima, Akira; Sakaguchi, Takashi

1990-01-01

The uranium-adsorbing abilities of seven plant wastes were investigated. High abilities to adsorb uranium from non-saline water containing 10 mg dm -3 of uranium were observed with a number of plant wastes tested. However, with seawater supplemented with 10 mg dm -3 of uranium, similar results were found only with chestnut residues. When the plant wastes were immobilized with formaldehyde, their ability to adsorb uranium was increased. Uranium and copper ions were more readily adsorbed by all plant wastes tested than other metal ions from a solution containing a mixture of seven different heavy metals. The selective adsorption of heavy metal ions differs with different species of plant wastes. The immobilization of peanut inner skin, orange peel and grapefruit peel increased the selectivity for uranium. (author)

Method of processing liquid wastes containing radioactive materials

International Nuclear Information System (INIS)

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

1983-01-01

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

Handling and storage of high-level liquid wastes from reprocessing of spent fuel

International Nuclear Information System (INIS)

Finsterwalder, L.

1982-01-01

The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

Treatment of radioactive wastes containing plutonium

International Nuclear Information System (INIS)

Orlando, O.S.; Aparicio, G.; Greco, L.; Orosco, E.H.; Cassaniti, P.; Salguero, D.; Toubes, B.; Perez, A.E.; Menghini, J.E.; Esteban, A.; Adelfang, P.

1987-01-01

The radioactive wastes generated in the process of manufacture and control of experimental fuel rods of mixed oxides, (U,Pu)O 2 , require an specific treatment due to the plutonium content. The composition of liquid wastes, mostly arising from chemical checks, is variable. The salt content, the acidity, and the plutonium and uranium content are different, which makes necessary a chemical treatment before the inclusion in concrete. The solid waste, such as neoprene gloves, PVC sleeves, filter paper, disposable or broken laboratory material, etc. are also included in concrete. In this report the methods used to dispose of wastes at Alpha Facility are described. With regard to the liquid wastes, the glove box built to process them is detailed, as well as the applied chemical treatment, including neutralization, filtration and later solidification. As for the solid wastes, it is described the cementation method consisting in introducing them into an expanded metal matrix, of the basket type, that contains as a concentric drum of 200 liter capacity which is smaller than the matrix, and the filling with wet cement mortar. (Author)

Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano oxide ferromagnetic

International Nuclear Information System (INIS)

Vuong Huu Anh; Nguyen Van Chinh; Nguyen Ba Tien; Doan Thi Thu Hien; Luu Cao Nguyen

2015-01-01

Nano oxide ferromagnetic Fe_3O_4 KT which was produced by the Military Institute of Science and Technology were used to adsorbed heavy metal elements in liquid waste. In this report, the nano oxide ferromagnetic Fe_3O_4 KT with the particle size of 80-100 nm and the specific surface area of 50-70 m"2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution were investigated. The results showed that the maximum adsorption capacity for uranium of the nano Fe_3O_4 KT was 53.5 mgU/g with conditions such as: room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours . From the results obtained, nano Fe_3O_4 KT was tested to treatment real liquid waste of uranium ore processing after removing almost heavy metals and a part of radioactive elements by preliminary precipitation at pH 8. The results were analyzed on the ICP-MS and α, β total activity equipment, the solution concentration after treatment suitable for Vietnamese Technical Regulation on industrial wastewater QCVN 40: 2011 (concentrations of heavy metals; total activity of α and β). (author)

Containment of solidified liquid hazardous waste in domal salt

International Nuclear Information System (INIS)

Domenico, P.A.; Lerman, A.

1992-01-01

In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

On the technical development to minimize the quantity of solid wastes in a uranium conversion

International Nuclear Information System (INIS)

Otomura, Keiichiro; Ogura, Yoshikazu; Fujisaki, Sakae

1987-01-01

We have developed the new process of treating the waste liquor from a uranium conversion at Ningyo Toge Works PNC, Japan. This process consists of neutralizing precipitation, solid liquid separation, distillation and adsorption. At a neutralizing precipitation step a magnesium oxide is added in the waste liquor containing uranium and fluorine. Most of the uranium and fluorine in the waste liquor precipitate as magnesium compounds. A sulfuric acid is added to the precipitate separated by a filter to dissolve. The resulting solution is then distilled to recover a hydrofluoric acid as a distillate. Uranium is recovered from a residue by an anion exchange method. The recoverd fluorine and uranium are recycled to the main process of conversion. The filtrate separated at the precipitation step is then passed through adsorbing columns. The residual fluorine and uranium in the filtrate were adsorbed and removed by the chelating resine which selectively adsorb the uranium and fluorine. After that the treated waste liquor is discharged out of the plant. This process has merits of being able to minimize the quantity of solid waste in comparison with the conventional process and to recover uranium and fluorine. This process can also be applied to uranium reconversion process from uranium hexafluoride to uranium oxide and to uranium metal production process, which produce the same kind of waste liquor. (author)

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Use of diatomaceous to liquid organic wastes adsorption

International Nuclear Information System (INIS)

Sanhueza M, Azucena; Padilla S, Ulises

1999-01-01

Background: One of the radioactive wastes that the Radioactive Wastes Management Unit must process are organic liquids from external generators and from sections of the Chilean Nuclear Energy Commission (CCHEN). The wastes from external generators contain H 3 and C 14; while the wastes from the CCHEN are contaminated with uranium. The total volume of liquid organic wastes that must be treated is 5 m3. The options recommended for processing these wastes are incineration or the adsorption of the organic liquid by some adsorbing medium and its subsequent immobilization in cement molds. Due to the cost of incineration, the adsorption method was chosen for study. Objective: To find the optimum amount of adsorbent to be saturated with radioactive organic liquid from liquid scintillation and to study immobilization in cement molds. Methodology: Adsorption granulated (1568 Merck) and diatom earth were tested as adsorbent mediums. The adsorbents were mixed in different ratios of volume with the organic liquid. Then the waste was mixed with different water/cement ratios to define the best immobilization conditions. Conclusions: The tests carried out with 2 adsorbents recommended in the literature and available in the CCHEN show that as adsorbent waste ratio decreases, the percentage of liquid adsorbed increases, as expected: a greater volume of adsorbent retains a greater quantity of liquid, with an increase in the final volume, depending on the adsorbent used. Of these adsorbents, the diatom earth was better for treating liquid organic wastes. It had 100% adsorption and an increased volume of 0%, which is more than enough from the volumetric point of view of waste management. The ratio 0.8 liquid/adsorbent also showed good characteristics, but more study is needed to decide on the above, since liquid remains to be adsorbed. This work must continue to study the repeatability of results, to obtain physical and radiological characteristics for the immobilized products and to

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

Energy Technology Data Exchange (ETDEWEB)

Adamovich, D. [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation); Batorshin, G.; Logunov, M.; Musalnikov, A. [FSUE ' PO Mayak' , 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)

2013-07-01

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

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid

International Nuclear Information System (INIS)

Sung, J.; Kim, J.; Lee, Y.; Seol, J.; Ryu, J.; Park, K.

2011-01-01

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 deg. C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined. (authors)

Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano Fe_3O_4 KT particles

International Nuclear Information System (INIS)

Vuong Huu Anh; Nguyen Ba Tien; Doan Thi Thu Hien; Luu Cao Nguyen; Nguyen Van Chinh

2015-01-01

Nano Fe_3O_4 KT was produced from the Military Institute of Science and Technology were used to adsorbed heavy metal elements in liquid waste. In this report, the nano Fe_3O_4 KT particles sized 80-100 nm and specific surface area was 50-70 m"2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution. The results showed the maximum adsorption capacity of the nano Fe_3O_4 KT was 53.5 mg/g with conditions such as room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours (Uranium/materials). From the results obtained, nano Fe_3O_4 KT tested to treatment liquid waste of uranium ore processing after preliminary precipitation removed almost heavy metals and a part of radioactive elements. The results were analyzed on the ICP-MS and α, β total counting, instrument. The solution concentration after treatment was suitable for Vietnam discharge standards into environment (QCVN 40:2011 on Industrial wastewater). (author)

Feasibility studies on electrochemical recovery of uranium from solid wastes contaminated with uranium using 1-butyl-3-methylimidazorium chloride as an electrolyte

Energy Technology Data Exchange (ETDEWEB)

Ohashi, Yusuke, E-mail: ohhashi.yusuke@jaea.go.jp [Ningyo-toge Environmental Engineering center, Japan Atomic Energy Agency, 1550 Kamisaibara, Kagamino-cho, Tomata-gun, Okayama 708-0698 (Japan); Harada, Masayuki [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-34 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Asanuma, Noriko [Department of Nuclear Engineering, School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Ikeda, Yasuhisa [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-34 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2015-09-15

Highlights: • The uranium component of steel wastes and spent NaF adsorbent are easily dissolved into BMICl. • The uranyl(VI) species in BMICl are reduced to U(V) irreversibly around −0.8 to −1.3 V. • The dissolved uranium species in BMICl are recovered as black deposits electrolytically. • The deposit is the mixtures of U(IV) and U(VI) compounds containing O, F, Cl, and N elements. - Abstract: In order to examine feasibility of the electrochemical deposition method for recovering uranium from the solid wastes contaminated with uranium using ionic liquid as electrolyte, we have studied the electrochemical behavior of each solution prepared by soaking the spent NaF adsorbents and the steel waste contaminated with uranium in BMICl (1-butyl-3-methyl- imidazolium chloride). The uranyl(VI) species in BMICl solutions were found to be reduced to U(V) irreversibly around −0.8 to −1.3 V vs. Ag/AgCl. The resulting U(V) species is followed by disproportionation to U(VI) and U(IV). Based on the electrochemical data, we have performed potential controlled electrolysis of each solution prepared by soaking the spent NaF adsorbents and steel wastes in BMICl at −1.5 V vs. Ag/AgCl. Black deposit was obtained, and their composition analyses suggest that the deposit is the mixtures of U(IV) and U(VI) compounds containing O, F, Cl, and N elements. From the present study, it is expected that the solid wastes contaminated with uranium can be decontaminated by treating them in BMICl and the dissolved uranium species are recovered electrolytically.

Treatment of liquid effluents from uranium analytical method 'DAVIES & GRAY' by electrodialysis and electrodialysis reactive tests

International Nuclear Information System (INIS)

Zuniga Alvear, Karina Andrea

2014-01-01

This work describes the process which produces liquid waste coming from the chemical analysis laboratory of the Chilean Nuclear Energy Commission (CCHEN), from the analytical technique called 'Davies and Gray' and their further treatment, using electro dialysis (ED) and reactive electro dialysis (RED), in order to achieve lower uranium contents in solution. The contamination in water is a big problem, since there are many places in the world where is limited. For these reasons new treatments must be done, and the ion-selective membrane has opened a new path for these processes. The radioactive liquid waste have lots of other restrictions in their final disposal, which difficult even more the water recovery, because the law has very strict security margins with respect to these ones. In the case of liquid waste containing uranium, the concern increases, because being the uranium a radioactive element has it has to be lowered at its maximum, or eliminated directly, in order to avoid any kind of contamination. There exist national regulations and international recommendations. They have stipulated the correct management and disposal for radioactive waste. These can come from any uranium production process. In any of these, the liquid waste contains certain uranium content, which after the end of the process; the discarded waste must go through a conditioning and cleaning process for its afterward liberation or recycling. In this study, it was tested the electro dialysis as a radioactive waste treatment, only uranium containing waste coming from the chemical analysis laboratory in CCHEN. The electro dialysis process has a direct competition with other separation process, such as distillation, ionic exchange, and reverse osmosis, among others. The classic electro dialysis has been developed during the 50's, and until today, there has been different version, as inverse, reactive, reversible. The unidirectional and reactive electro dialysis will be the

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)

Application of macrophytes as biosorbents for radioactive liquid waste treatment

International Nuclear Information System (INIS)

Vieira, Ludmila Cabreira

2016-01-01

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

Sulphate in Liquid Nuclear Waste: from Production to Containment

Energy Technology Data Exchange (ETDEWEB)

Lenoir, M.; Grandjean, A.; Ledieu, A.; Dussossoy, J.L.; Cau Dit Coumes, C.; Barre, Y.; Tronche, E. [CEA Marcoule, DEN/DTCD/SECM/LDMC, Batiment 208 BP17171, Bagnols sur Ceze, 30207 (France)

2009-06-15

Nuclear industry produces a wide range of low and intermediate level liquid radioactive wastes which can include different radionuclides such as {sup 90}Sr. In La Hague reprocessing plant and in the nuclear research centers of CEA (Commissariat a l'Energie Atomique), the coprecipitation of strontium with barium sulphate is the technique used to treat selectively these contaminated streams with the best efficiency. After the decontamination process, low and intermediate level activity wastes incorporating significant quantities of sulphate are obtained. The challenge is to find a matrix easy to form and with a good chemical durability which is able to confine this kind of nuclear waste. The current process used to contain sulphate-rich nuclear wastes is bituminization. However, in order to improve properties of containment matrices and simplify the process, CEA has chosen to supervise researches on other materials such as cements or glasses. Indeed, cements are widely used for the immobilization of a variety of wastes (low and intermediate level wastes) and they may be an alternative matrix to bitumen. Even if Portland cement, which is extensively used in the nuclear industry, presents some disadvantages for the containment of sulphate-rich nuclear wastes (risk of swelling and cracking due to delayed ettringite formation), other cement systems, such as calcium sulfo-aluminate binders, may be valuable candidates. Another matrix to confine sulphate-rich waste could be the glass. One of the advantages of this material is that it could also immobilize sulphate containing high level nuclear waste which is present in some countries. This waste comes from the use of ferrous sulfamate as a reducing agent for the conversion of Pu{sup 4+} to Pu{sup 3+} in the partitioning stage of the actinides during reprocessing. Sulphate solubility in borosilicate glasses has already been studied in CEA at laboratory and pilot scales. At a pilot scale, low level liquid waste has been

Evaluation of nanofiltration membranes for treatment of liquid radioactive waste

International Nuclear Information System (INIS)

Oliveira, Elizabeth Eugenio de Mello

2013-01-01

The physicochemical behavior of two nanofiltration membranes for treatment of a low-level radioactive liquid waste (carbonated water) was investigated through static, dynamic and concentration tests. This waste was produced during conversion of uranium hexafluoride (UF 6 ) to uranium dioxide (UO 2 ) in the cycle of nuclear fuel. This waste contains about 7.0 mg L -1 of uranium and cannot be discarded to the environment without an adequate treatment. In static tests membrane samples were immersed in the waste for 24 to 5000 h. Their transport properties (hydraulic permeability, permeate flux, sulfate and chloride ions rejection) were evaluated before and after immersion in the waste using a permeation flux front system under 0.5 MPa. The selective layer (polyamide) was characterized by zeta potential, contact angle, scanning electron microscopy for field emission, atomic force microscopy, infrared spectroscopy, x-ray fluorescence and thermogravimetric analysis before and after static tests. In dynamic tests the waste was permeated under 0.5 MPa, and the membranes showed rejection to uranium above 85% were obtained. The short-term static tests (24-72 h) showed that the selective layer and surface charge of the membranes were not chemical changed, according infrared spectra data. After 5000 h a coating layer was released from the membranes, poly(vinyl alcohol), PVA. After this loss the rejection for uranium decreased. Permeation and concentration of the waste were carried out in permeation flux tangential system under 1.5 MPa. The rejection of uranium was around 90% for permeation tests. In concentration tests the permeated was collected continuously until about 80% reduction of the feed volume. The rejection of uranium was of the 97%. The nanofiltration membranes tested were efficient to concentrate the uranium from the waste. (author)

Liquid membranes and process for uranium recovery therewith

International Nuclear Information System (INIS)

Frankenfeld, J.W.; Li, N.N.T.; Bruncati, R.L.

1981-01-01

A liquid membrane system consisting of water-in-oil type emulsions dispersed in water, which is capable of extracting uranium-containing ions from an aqueous feed solution containing uranium ions at a temperature in the range of 25 0 C to 80 0 C, is described. The emulsion comprises an aqueous interior phase surrounded by a surfactant-containing exterior phase. The exterior phase is immiscible with the interior phase and comprises a transfer agent capable of transporting selectively the desired uranium-containing ions and a solvent for the transfer agent. The interior phase comprises a reactant capable of removing uranium-containing ions from the transfer agent and capable of changing the valency of the uranium in uranium-containing ions to a second valency state and converting the uranium-containing ions into a nonpermeable form. (U.K.)

Treatment of uranium-containing effluent in the process of metallic uranium parts

International Nuclear Information System (INIS)

Yuan Guoqi

1993-01-01

The anion exchange method used in treatment of uranium-containing effluent in the process of metallic parts is the subject of the paper. The results of the experiments shows that the uranium concentration in created water remains is less than 10 μg/l when the waste water flowed through 10000 column volume. A small facility with column volume 150 litre was installed and 1500 m 3 of waste water can be cleaned per year. (1 tab.)

Treatment of radioactive liquid organic waste using bacteria community

International Nuclear Information System (INIS)

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

2012-01-01

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

Uranium dioxide in Fe(III)-containing ionic liquids with DMSO: Dissolution, separation, and structural characterization

Energy Technology Data Exchange (ETDEWEB)

Yao, Aining; Chu, Taiwei, E-mail: twchu@pku.edu.cn

2016-11-15

UO{sub 2} can be successfully dissolved in imidazolium-based Fe(III)-containing ionic liquids (ILs) with the help of DMSO. Spectroscopic studies and X-ray diffraction show that UO{sub 2}Cl{sub 4}{sup 2−} is the principal product. The dissolved uranyl species can be easily separated from the ILs via a combination of crystallization and solvent extraction. Moreover, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd, compared with the total amount of uranium and the rare-earth elements, exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. The solvents of acetone and acetonitrile could be used to separate the rare-earth elements from uranium in the IL with the help of imidazolium chloride. Considering the complete process from the dissolution of UO{sub 2} and some rare-earth oxides to the separation of uranium and rare-earth elements in the IL, the facile approach is promising for the spent nuclear fuel reprocessing. - Graphical abstract: UO{sub 2} can be successfully dissolved in Fe-containing ILs with the help of DMSO to form UO{sub 2}Cl{sub 4}{sup 2−}. The rare earth elements of Sm, Eu, and Gd can be separated from uranium in the IL, and meanwhile, the recovery of dissolved uranyl species and Fe-containing IL can also be achieved. - Highlights: • Dissolution of UO{sub 2} can be successfully achieved in imidazolium-based Fe-containing ILs with the help of DMSO without additional oxidants. • Compared with the total amount of uranium and the rare-earth elements, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. • The separation of the rare-earth elements from uranium has also been achieved via a combination of crystallization and solvent extraction.

Uranium dioxide in Fe(III)-containing ionic liquids with DMSO: Dissolution, separation, and structural characterization

International Nuclear Information System (INIS)

Yao, Aining; Chu, Taiwei

2016-01-01

UO_2 can be successfully dissolved in imidazolium-based Fe(III)-containing ionic liquids (ILs) with the help of DMSO. Spectroscopic studies and X-ray diffraction show that UO_2Cl_4"2"− is the principal product. The dissolved uranyl species can be easily separated from the ILs via a combination of crystallization and solvent extraction. Moreover, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd, compared with the total amount of uranium and the rare-earth elements, exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. The solvents of acetone and acetonitrile could be used to separate the rare-earth elements from uranium in the IL with the help of imidazolium chloride. Considering the complete process from the dissolution of UO_2 and some rare-earth oxides to the separation of uranium and rare-earth elements in the IL, the facile approach is promising for the spent nuclear fuel reprocessing. - Graphical abstract: UO_2 can be successfully dissolved in Fe-containing ILs with the help of DMSO to form UO_2Cl_4"2"−. The rare earth elements of Sm, Eu, and Gd can be separated from uranium in the IL, and meanwhile, the recovery of dissolved uranyl species and Fe-containing IL can also be achieved. - Highlights: • Dissolution of UO_2 can be successfully achieved in imidazolium-based Fe-containing ILs with the help of DMSO without additional oxidants. • Compared with the total amount of uranium and the rare-earth elements, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. • The separation of the rare-earth elements from uranium has also been achieved via a combination of crystallization and solvent extraction.

Effect of liquid waste discharges from steam generating facilities

Energy Technology Data Exchange (ETDEWEB)

McGuire, H.E. Jr.

1977-09-01

This report contains a summary of the effects of liquid waste discharges from steam electric generating facilities on the environment. Also included is a simplified model for use in approximately determining the effects of these discharges. Four basic fuels are used in steam electric power plants: three fossil fuels--coal, natural gas, and oil; and uranium--presently the basic fuel of nuclear power. Coal and uranium are expected to be the major fuels in future years. The following power plant effluents are considered: heat, chlorine, copper, total dissolved solids, suspended solids, pH, oil and grease, iron, zinc, chrome, phosphorus, and trace radionuclides.

Effect of liquid waste discharges from steam generating facilities

International Nuclear Information System (INIS)

McGuire, H.E. Jr.

1977-09-01

This report contains a summary of the effects of liquid waste discharges from steam electric generating facilities on the environment. Also included is a simplified model for use in approximately determining the effects of these discharges. Four basic fuels are used in steam electric power plants: three fossil fuels--coal, natural gas, and oil; and uranium--presently the basic fuel of nuclear power. Coal and uranium are expected to be the major fuels in future years. The following power plant effluents are considered: heat, chlorine, copper, total dissolved solids, suspended solids, pH, oil and grease, iron, zinc, chrome, phosphorus, and trace radionuclides

Liquid uranium contaimment in refractories metals

International Nuclear Information System (INIS)

Duarte, J.L.; Padilha, A.F.

1982-01-01

Tests were performed on metalic materials for liquid uranium containment up to 2100 0 C. The materials Nb, Mo, Ta and W in the form of crucibles were tested at 2100 0 C for one hour in the presence of flowing argon. After testing, the crucibles were etched using HCl and analysed by optical metallography and electron proble microanalysis. The results are discussed in terms of Berthoud equation and indicated that the solubility limit of the crucible material in uranium at the temperature controlls the crucible dissolution by liquid uranium. The various phases formed, the mechanism of dissolution and the possible material for future use are presented and discussed. (Author) [pt

Liquid membrane process for uranium recovery

International Nuclear Information System (INIS)

Valint, P.L. Jr.

1982-01-01

An improved liquid membrane emulsion extraction process for recovering uranium from a WPPA feed solution containing uranyl cations wherein said feed is contacted with a water-in-oil emulsion which extracts and captures the uranium in the interior aqueous phase thereof, wherein the improvement comprises the presence of an alkane diphosphonic acid uranium complexing agent in the interior phase of the emulsion. This improvement results in greater extraction efficiency

Biochemical process of low level radioactive liquid simulation waste containing detergent

International Nuclear Information System (INIS)

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

2015-01-01

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

Biochemical process of low level radioactive liquid simulation waste containing detergent

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-29

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

Biochemical process of low level radioactive liquid simulation waste containing detergent

Science.gov (United States)

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

2015-12-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Can ionic liquids be used as templating agents for controlled design of uranium-containing nanomaterials?

International Nuclear Information System (INIS)

Visser, Ann E.; Bridges, Nicholas J.; Tosten, Michael H.

2013-01-01

Graphical abstract: - Highlights: • Uranium oxides nanoparticles prepared using ionic liquids. • IL cation alkyl length impacts oxide morphology. • Low temperature UO 2 synthesis. - Abstract: Nanostructured uranium oxides have been prepared in ionic liquids as templating agents. Using the ionic liquids as reaction media for inorganic nanomaterials takes advantage of the pre-organized structure of the ionic liquids which in turn controls the morphology of the inorganic nanomaterials. Variation of ionic liquid cation structure was investigated to determine the impact on the uranium oxide morphologies. For two ionic liquid cations, increasing the alkyl chain length increases the aspect ratio of the resulting nanostructured oxides. Understanding the resulting metal oxide morphologies could enhance fuel stability and design

Fundamental study on decontamination of wastes contaminated by uranium fluorides by using ionic liquids - dissolution and electrochemistry of uranium in 1-butyl-3-methylimidazolium chloride

International Nuclear Information System (INIS)

Noriko Asanuma; Yusuke Ohhashi; Yukio Wada; Masayuki Harada; Yasuhisa Ikeda

2008-01-01

Treatment method for wastes contaminated uranium fluorides by using ionic liquids as media of pyrochemical process instead of alkali metal chloride molten salts was proposed. In this method, uranium fluorides such as UF 4 or NaF adsorbing UF 6 are dissolved in 1-butyl-3-methylimidazolium chloride (BMICl) and dissolved uranium species are recovered as deposits by electrochemical reduction. Under the atmospheric condition, UF 4 was completely dissolved in BMICl at 100 deg. C. UV-vis absorption spectra of the sample solution indicated that main species of uranium are U(VI) and a part of uranium exists as U(IV). Chemical form of uranium in the NaF adsorbents is Na 3 UO 2 F 5 . Therefore, it was immediately dissolved to BMICl. However, complete dissolution was not achieved. Cyclic voltammetry of the solutions prepared by dissolution experiments was performed. Redox properties of uranium species in each sample were irreversible. It was assigned to reduction of U(VI) to U(IV). As a result of preliminary bulk electrolysis, it was expected that reduction products are deposited on the carbon cathode. (authors)

Method of processing radioactive liquid wastes

International Nuclear Information System (INIS)

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

1984-01-01

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

The utilization of uranium industry technology and relevant chemistry to leach uranium from mixed-waste solids

International Nuclear Information System (INIS)

Mattus, A.J.; Farr, L.L.

1991-01-01

Methods for the chemical extraction of uranium from a number of refractory uranium-containing minerals found in nature have been in place and employed by the uranium mining and milling industry for nearly half a century. These same methods, in conjunction with the principles of relevant uranium chemistry, have been employed at the Oak Ridge National Laboratory (ORNL) to chemically leach depleted uranium from mixed-waste sludge and soil. The removal of uranium from what is now classified as mixed waste may result in the reclassification of the waste as hazardous, which may then be delisted. The delisted waste might eventually be disposed of in commercial landfill sites. This paper generally discusses the application of chemical extractive methods to remove depleted uranium from a biodenitrification sludge and a storm sewer soil sediment from the Y-12 weapons plant in Oak Ridge. Some select data obtained from scoping leach tests on these materials are presented along with associated limitations and observations which might be useful to others performing such test work. 6 refs., 2 tabs

The utilization of uranium industry technology and relevant chemistry to leach uranium from mixed-waste solids

Energy Technology Data Exchange (ETDEWEB)

Mattus, A.J.; Farr, L.L.

1991-01-01

Methods for the chemical extraction of uranium from a number of refractory uranium-containing minerals found in nature have been in place and employed by the uranium mining and milling industry for nearly half a century. These same methods, in conjunction with the principles of relevant uranium chemistry, have been employed at the Oak Ridge National Laboratory (ORNL) to chemically leach depleted uranium from mixed-waste sludge and soil. The removal of uranium from what is now classified as mixed waste may result in the reclassification of the waste as hazardous, which may then be delisted. The delisted waste might eventually be disposed of in commercial landfill sites. This paper generally discusses the application of chemical extractive methods to remove depleted uranium from a biodenitrification sludge and a storm sewer soil sediment from the Y-12 weapons plant in Oak Ridge. Some select data obtained from scoping leach tests on these materials are presented along with associated limitations and observations which might be useful to others performing such test work. 6 refs., 2 tabs.

Method for converting uranium oxides to uranium metal

Science.gov (United States)

Duerksen, Walter K.

1988-01-01

A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

Assessment of nonpoint source chemical loading potential to watersheds containing uranium waste dumps associated with uranium exploration and mining, Browns Hole, Utah

Science.gov (United States)

Marston, Thomas M.; Beisner, Kimberly R.; Naftz, David L.; Snyder, Terry

2012-01-01

During August of 2008, 35 solid-phase samples were collected from abandoned uranium waste dumps, undisturbed geologic background sites, and adjacent streambeds in Browns Hole in southeastern Utah. The objectives of this sampling program were (1) to assess impacts on human health due to exposure to radium, uranium, and thorium during recreational activities on and around uranium waste dumps on Bureau of Land Management lands; (2) to compare concentrations of trace elements associated with mine waste dumps to natural background concentrations; (3) to assess the nonpoint source chemical loading potential to ephemeral and perennial watersheds from uranium waste dumps; and (4) to assess contamination from waste dumps to the local perennial stream water in Muleshoe Creek. Uranium waste dump samples were collected using solid-phase sampling protocols. Solid samples were digested and analyzed for major and trace elements. Analytical values for radium and uranium in digested samples were compared to multiple soil screening levels developed from annual dosage calculations in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act's minimum cleanup guidelines for uranium waste sites. Three occupancy durations for sites were considered: 4.6 days per year, 7.0 days per year, and 14.0 days per year. None of the sites exceeded the radium soil screening level of 96 picocuries per gram, corresponding to a 4.6 days per year exposure. Two sites exceeded the radium soil screening level of 66 picocuries per gram, corresponding to a 7.0 days per year exposure. Seven sites exceeded the radium soil screening level of 33 picocuries per gram, corresponding to a 14.0 days per year exposure. A perennial stream that flows next to the toe of a uranium waste dump was sampled, analyzed for major and trace elements, and compared with existing aquatic-life and drinking-water-quality standards. None of the water-quality standards were exceeded in the stream samples.

Application of acid dissolution and natural evaporation to wet cake containing uranium

International Nuclear Information System (INIS)

Kim, Kil J.; Kang, Il Sik; Shon, Jong S.; Hong, Kwon P.

2005-01-01

Chemical wastes containing small amounts of uranium cause environmental problems, if those wastes exceed the concentration of the EPA standard, 20 μg.. /L, and the concentrated sludge should be additionally dried and packaged into a drum, and categorized as a radioactive waste. Diphosil resin is developed to specifically remove actinides or multivalent metals. The immobilization technique is adopted to make a bead form of Diphosil by embedding into sodium alginate, and adsorption characteristics for uranium are reported for a simulated waste solution. In this study, acid dissolution is applied to dissolve uranium from the precipitates of sludge or the dewatered cake in the reduced volume of wastes solution, and removal characteristics of uranium is experimented. From the results, the most effective treatment method for the dissolved solution is suggested

Uranium-bearing wastes and their radon emanation

International Nuclear Information System (INIS)

Sasaki, Tomozo; Imamura, Mitsutaka; Gunji, Yasuyoshi

2007-01-01

There are no data available with regard to radon emanation coefficients for uranium-bearing wastes; such data are needed for the assessment of radiation exposure from radon that will be generated in the distant future as one uranium progeny at shallow land disposal sites for uranium-bearing wastes. There are many kinds of uranium-bearing wastes. However, it is not necessary to measure the radon emanation coefficients for all of them for two reasons. First, the radon emanation coefficients for uranium-bearing wastes contaminated by dissolved uranium are determined by the uranium chemical form, the manner of uranium deposition on the waste matrix, and the size of the particles which constitute the waste matrix. Therefore, only a few representative measurements are sufficient for such uranium-bearing wastes. Second, it is possible to make theoretical calculations of radon emanation coefficients for uranium-bearing wastes contaminated by UO 2 particles before sintering. In the present study, simulated uranium-bearing wastes contaminated by dissolved uranium were prepared, their radon emanation coefficients were measured and radon emanation coefficients were calculated theoretically for uranium-bearing wastes contaminated by UO 2 particles before sintering. The obtained radon emanation coefficients are distributed at higher values than those for ubiquitous soils and rocks in the natural environment. Therefore, it is not correct to just compare uranium concentrations among uranium-bearing wastes, ubiquitous soils and rocks in terms of radiation exposure. The radon emanation coefficients obtained in the present study have to be employed together with the uranium concentration in uranium-bearing wastes in order to achieve proper assessment of radiation exposure. (author)

Fundamental study on decontamination of wastes contaminated by uranium fluorides by using ionic liquids - dissolution and electrochemistry of uranium in 1-butyl-3-methylimidazolium chloride

Energy Technology Data Exchange (ETDEWEB)

Noriko Asanuma [Department of Energy Science and Engineering, School of Engineering, Tokai University 1117 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Yusuke Ohhashi; Yukio Wada [Ningyo-toge Environmental Engineering Center, Japan Atomic Energy Agency Kagamino-cho, Tomata-gun, Okayama 708-0698 (Japan); Masayuki Harada; Yasuhisa Ikeda [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-34 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2008-07-01

Treatment method for wastes contaminated uranium fluorides by using ionic liquids as media of pyrochemical process instead of alkali metal chloride molten salts was proposed. In this method, uranium fluorides such as UF{sub 4} or NaF adsorbing UF{sub 6} are dissolved in 1-butyl-3-methylimidazolium chloride (BMICl) and dissolved uranium species are recovered as deposits by electrochemical reduction. Under the atmospheric condition, UF{sub 4} was completely dissolved in BMICl at 100 deg. C. UV-vis absorption spectra of the sample solution indicated that main species of uranium are U(VI) and a part of uranium exists as U(IV). Chemical form of uranium in the NaF adsorbents is Na{sub 3}UO{sub 2}F{sub 5}. Therefore, it was immediately dissolved to BMICl. However, complete dissolution was not achieved. Cyclic voltammetry of the solutions prepared by dissolution experiments was performed. Redox properties of uranium species in each sample were irreversible. It was assigned to reduction of U(VI) to U(IV). As a result of preliminary bulk electrolysis, it was expected that reduction products are deposited on the carbon cathode. (authors)

Method of conversion of caustic liquid radioactive wastes containing sodium nitrates into solid insoluble products

International Nuclear Information System (INIS)

Barney, G.S.; Brownell, L.E.

1975-01-01

A proposal is made to convert caustic, liquid, radioactive wastes containing sodium nitrate into a solid product by reaction with powdered aluminium silicate at temperatures between 30 0 and 100 0 C, which is practically insoluble (10 -7 to 10 -10 g/cm 2 -day) and is thermally stable. A cancrinite is formed which binds the radioactive salts in the cage-like structure of its crystal lattice. The method is also suitable for liquid wastes from the Purex method as well as for wastes containing fission products of Cs 137 and Sr 90 in concentrations of 0.37 M to 0.01 M. Numerous detailed examples explain the invention. (UW/LH) [de

Method of separation of uranium from contaminating ions in an aqueous feed liquid containing uranyl ions

International Nuclear Information System (INIS)

Sundar, P.S.; Elikan, L.; Lyon, W.L.

1975-01-01

A coupled cationic/anionic method for the separation of uranium from contaminated aqueous solutions which contain uranyl ions is proposed. The fluid is extracted using an organic solvent containing a reagent which, together with the uranyl ions, forms a soluble aggregate in that solvent. As an example, 0.1 - 1 Mol/l Di-2-ethyl-hexyl-phosphorous acid in kerosene is mentioned. The organic solvent is then treated with a sealing liquid (volume ratio 20 - 35). For separation, an aqueous carbonate solution or a sulfuric acid solution can be used; the most favorable pH-values and concentrations for both cases are mentioned. The U +4 -ion at the sulfuric acid separation is subsequently oxidized to the uranyl ion with air. In each case, an extraction with an amine follows; after that, the amine is separated using an ammonium-carbonate solution and the uranium aggregate is precipitated, for example as ammonium uranyl tricarbonate, and then further processed to uranium oxide. The solvents and fluids used are led back in closed circuit; a flow diagram is given. (UWI) [de

New system for the container conditioning of liquid waste in the German future finale repository 'Schacht Konrad'

International Nuclear Information System (INIS)

Starke, H.

2012-01-01

The full text of publication follows. On-site the NPP Gundremmingen liquid radioactive waste from the NPP water treatment plant is stored in resin or concentrate collecting tanks. These liquid wastes are cemented in containers in order to temporarily store them in the Bavarian interim storage Mitterteich until they are transported into final repository in 'Schacht Konrad'. With this new system liquid radioactive waste is for the first time conditioned directly into containers destined for final repository in 'Schacht Konrad'. Thus, a very secure and sustainable procedure was developed which also provides high profitability. The conditioning plant for resins and concentrate extracts the liquid waste from the respective collecting tank and transports the waste to the separation tank. This separation tank is dimensioned to ensure complete filling of a Konrad container with only one batch. Within the tank there is the option to adjust the suspensions solids content by either extracting supernatant water or by adding de-ionised water. The specific activity is analysed and after the radiologic data and the solids content are available, the containers are cemented. The required amount of cement is based on the solids content and is automatically added. In the mixer, cement and primary waste suspension are mixed. This mixture is filled into the Konrad container via the allocator. The allocator is a funnel-shaped inlet equipped with a movable tube which makes sure the mixture is evenly spread and also ensures optimal filling of the Konrad container. While filling is ongoing, the container is covered by a lowerable splash guard to avoid contamination. The room situation in Gundremmingen and the specific activities of the primary waste suspension make it necessary to disperse the plant to several rooms. Main components such as separation tanks and pumps are installed in shielded rooms. All activities are conducted remotely controlled and are supervised from the central

Microbial transformation of uranium in wastes

International Nuclear Information System (INIS)

Francis, A.J.; Dodge, C.J.; Gillow, J.B.; Cline, J.E.

1989-01-01

Contamination of soils, water, and sediments by radionuclides and toxic metals from the disposal of uranium processing wastes is a major national concern. Although much is known about the physico- chemical aspects of U, we have little information on the effects of aerobic and anaerobic microbial activities on the mobilization or immobilization of U and other toxic metals in mixed wastes. In order to understand the mechanisms of microbial transformations of uranium, we examined a contaminated pond sediment and a sludge sample from the uranium processing facility at Y-12 Plant, Oak Ridge, TN. The uranium concentration in the sediment and sludge samples was 923 and 3080 ug/g dry wt, respectively. In addition to U, the sediment and sludge samples contained high levels of toxic metals such as Cd, Cr, Cu, Hg, Pb, Ni, and Zn. The association of uranium with the various mineral fractions of the sediment and sludge was determined by selective chemical extraction techniques. Uranium was associated to varying degrees with the exchangeable carbonate, iron oxide, organic, and inert fractions in both samples. Initial results in samples amended with carbon and nitrogen indicate immobilization of U due to enhanced indigenous microbial activity under anaerobic conditions. 23 refs., 4 figs., 5 tabs

Extraction of uranium(VI) by emulsion liquid membrane containing 5,8-diethyl-7-hydroxy-6-dodecanone oxime

International Nuclear Information System (INIS)

Akiba, Kenichi; Takahashi, Toshihiko; Kanno, Takuji

1984-01-01

Extraction of uranium(VI) by a liquid surfactant membrane has been studied. The stability of water-in-oil (w/o) emulsion dispersed in the continuous aqueous phase increased with an increase in surfactant concentrations and in the fraction of the organic phase in emulsion globules. Uranium(VI) in dilute acid solutions was extracted into (w/o) emulsions containing 5,8-diethyl-7-hydroxy-6-dodecanone oxime (LIX 63) as a mobile carrier and its concentration decreased according to [U]sub(t)=[U]sub(o)exp(-ksub(obsd)t). The apparent rate constants (ksub(obsd)) increased with an increase in carrier concentrations and in external pH values, while they were slightly dependent on the stripping acid concentrations. Uranium was transported and concentrated into the internal aqueous droplets. The final concentration of uranium in the external aqueous phase dropped to about 10 -3 of its initial value. (author)

Biodegradation of uranium-contaminated waste oil

International Nuclear Information System (INIS)

Hary, L.F.

1983-01-01

The Portsmouth Gaseous Diffusion Plant routinely generates quantities of uranium-contaminated waste oil. The current generation rate of waste oil is approximately 2000 gallons per year. The waste is presently biodegraded by landfarming on open field soil plots. However, due to the environmental concerns associated with this treatment process, studies were conducted to determine the optimum biodegradation conditions required for the destruction of this waste. Tests using respirometric flasks were conducted to determine the biodegradation rate for various types of Portsmouth waste oil. These tests were performed at three different loading rates, and on unfertilized and fertilized soil. Additional studies were conducted to evaluate the effectiveness of open field landfarming versus treatment at a greenhouse-like enclosure for the purpose of maintaining soil temperatures above ambient conditions. The respirometric tests concluded that the optimum waste oil loading rate is 10% weight of oil-carbon/weight of soil (30,600 gallons of uranium-contaminated waste oil/acre) on soils with adjusted carbon:nitrogen and carbon:phosphorus ratios of 60:1 and 800:1, respectively. Also, calculational results indicated that greenhouse technology does not provide a significant increase in biodegradation efficiency. Based on these study results, a 6300 ft. 2 abandoned anaerobic digester sludge drying bed is being modified into a permanent waste oil biodegradation facility. The advantage of using this area is that uranium contamination will be contained by the bed's existing leachate collection system. This modified facility will be capable of handling approximately 4500 gallons of waste oil per year; accordingly current waste generation quantities will be satisfactorily treated. 15 refs., 14 figs., 4 tabs

Recycling of Metal Containing Waste by Liquid-Liquid Extraction

International Nuclear Information System (INIS)

Reinhardt, H.

1999-01-01

Through the years, a large number of liquid-liquid extraction have been proposed for metal waste recovery and recycling(1,2). However, few of them have achieved commercial application. In fact, relatively little information is available on practical operation and economic feasibility. This presentation will give complementary information by describing and comparing three processes, based on the Am MAR hydrometallurgical concept and representing three different modes of operation

The uranium waste fluid processing examination by liquid and liquid extraction method using the emulsion flow method

International Nuclear Information System (INIS)

Kanda, Nobuhiro; Daiten, Masaki; Endo, Yuji; Yoshida, Hideaki; Mita, Yutaka; Naganawa, Hirochika; Nagano, Tetsushi; Yanase, Nobuyuki

2015-03-01

Spent centrifuges which had used for the development of the uranium enrichment technology are stored in the uranium enrichment facility located in Ningyo-toge Environmental Center, Japan Atomic Energy Agency (JAEA). Our technology of the centrifugal machine processing are supposed to separate the radioactive material adhered on surface of inner parts of centrifuges by the wet way decontamination method using the ultrasonic bath filled dilute sulfuric acid and water, and it is generated the neutralization sediment (sludge) by the processing of the radioactive waste fluid with the decontamination. JAEA had been considering the applicability of a streamlining and reduction of the processing of the sludge by decreases radioactive concentration including the sludge through the removes uranium from the radioactive waste fluid. As part of considerations, JAEA have been promoting technological developments of the uranium extraction separation using The Emulsion Flow Extraction Method (a theory propounded by JAEA-Nuclear Science and Engineering Center) in close coordination and cooperation between with JAEA-Nuclear Science and Engineering Center and Ningyo-toge Environmental Center from 2007 fiscal year. This report describes the outline of the application test using actual waste fluid of dilute sulfuric acid and water by developed the examination system introducing the emulsion flow extraction method. (author)

Processing method and device for radioactive waste containing surfactant

International Nuclear Information System (INIS)

Yukita, Atsushi; Yoshikawa, Ryozo; Izumida, Tatsuo; Nishi, Takashi; Hattori, Yasuo.

1997-01-01

Washing liquid wastes generated in washing facilities in a nuclear power plant are collected in a liquid waste collecting tank. A suspension containing a powdery active carbon is supplied to the liquid waste collecting tank. Organic ingredients such as of a surfactant, oil ingredients and radioactive materials in the form of ions contained in the washing liquid wastes are adsorbed to the powdery active carbon. The washing liquid wastes containing the powdery active carbon and granular radioactive materials are led into an active carbon separating and drying device. The powdery active carbon and granular radioactive materials contained in the washing liquid wastes are filtered and separated by a filtering plate, and accumulated as filtered materials on the surface of the filtering plate. The purified washing liquid wastes are discharged to the outside. The filtered materials are dried by hot steams (or hot water) and dried air. The filtered materials are peeled from the filtering plate. The filtered materials, in other word, dried powdery active carbon and granular radioactive materials are transported to and burnt in an incinerator. (I.N.)

Solidification of acidic liquid waste from 99Mo isotope production

International Nuclear Information System (INIS)

Parsons, G.J.

2001-01-01

Full text: The production of the radioisotope molybdenum-99 by the fission process began at ANSTO in the late 1960's. Molybdenum-99, with a half life of 66 hours, decays by beta emission to produce technetium-99m, a metastable isotope. Technetium-99m is the most widely used medical radioisotope due to its near ideal properties, particularly the radioactive half life of only 6 hours. ANSTO has been producing generators for around 30 years for distribution to hospitals and nuclear medicine centres. These generators produce technetium-99m for medical use by decay of the contained molybdenum-99. To produce molybdenum-99, uranium dioxide pellets enriched to 2.2% 235 U are irradiated in ANSTO's HIFAR reactor for about one week. The irradiated pellets are subsequently dissolved in nitric acid to allow the recovery of the molybdenum. An acidic intermediate level liquid waste results from this processing. A primary waste results from the raw leach solution (after removal of the molybdenum onto a packed alumina column) and a weaker secondary waste is produced from a series of column washing steps. The waste solution contains uranium, the majority of the other fission products and low levels of ammonia in a nitric acid solution. This liquid waste had been accumulating and stored in specially designed shielded tanks in a storage facility. A process has been developed at ANSTO to convert this intermediate level liquid waste into a crystalline solid form of considerably less volume and mass, for improved storage. The operation comprises three processing steps. The lower strength secondary waste solution first requires concentration, with the removal of water and some acid into a condensate. The condensate is chemically neutralised and treated through the conventional water treatment plant. Concentrated solution is then treated in a batch chemical process to reduce the low levels of ammonia to very low levels. The final evaporation process removes further water and acid and

Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

International Nuclear Information System (INIS)

Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

1998-05-01

To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

Design of chemical treatment unit for radioactive liquid wastes in Serpong nuclear facilities

International Nuclear Information System (INIS)

Salimin, Z.; Walman, E.; Santoso, P.; Purnomo, S.; Sugito; Suwardiyono; Wintono

1996-01-01

The chemical treatment unit for radioactive liquid wastes arising from nuclear fuel fabrication, radioisotopes production and radiometallurgy facility has been designed. The design of chemical processing unit is based on the characteristics of liquid wastes containing fluors from uranium fluoride conversion process to ammonium uranyl carbonate on the fuel fabrication. The chemical treatment has the following process steps: coagulation-precipitation of fluoride ion by calcium hydroxide coagulant, separation of supernatant solution from sludge, coagulation of remaining fluoride on the supernatant solution by alum, separation of supernatant from sludge, and than precipitation of fluors on the supernatant by polymer resin WWS 116. The processing unit is composed of 3 storage tanks for raw liquid wastes (capacity 1 m 3 per tank), 5 storage tanks for chemicals (capacity 0.5 m 3 per tank), 2 mixing reactors (capacity 0.5 m 3 per reactor), 1 storage tank for supernatant solution (capacity 1 m 3 ), and 1 storage tank for sludge (capacity 1 m 3 )

Recent activity on disposal of uranium waste

International Nuclear Information System (INIS)

Fujiwara, Noboru

1999-01-01

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

Radioactive waste management at Institute for Nuclear Research (ICN) - Pitesti

International Nuclear Information System (INIS)

Bujoreanu, C.

2004-01-01

The amounts of liquid and solid wastes accumulated at the Radioactive Wastes Treatment Plant are given. The technologies used for the treatment and conditioning of radioactive wastes are presented. The final product is metallic drum-concrete-radioactive wastes (type A package) for the final disposal at the National Repository Baita, Bihor. The facilities for radioactive waste management at ICN Pitesti are: Plant for treatment, with uranium recovery of liquid radioactive waste resulting from the fabrication of CANDU type nuclear fuel; Plant for treatment of low-active liquid wastes; Plant for conditioning in concrete of the radioactive concentrate obtained during the evaporation treatment of liquid radioactive waste; Plant for incineration of solid radioactive waste contaminated with natural uranium; Plant for treatment and conditioning of organic liquid radioactive waste with tritium content. This wastes are generated by Cernavoda-NPP operation; Plant for conditioning into bitumen of spent ion exchangers at TRIGA reactor. The existing Facility is Baita repository - with two rock cavities of an uranium mine and the total capacity of 21000 containers (200 l drums)

Remediation and upgrading of old, inadequate waste management facilities. Integrated waste management system for rare earth and rare metal industry at Sillamaee, Estonia, former uranium facility

International Nuclear Information System (INIS)

Kaasik, Tonis; Siinmaa, Anti

2001-01-01

Full text: The Sillamaee Metallurgical Plant was built in 1946-1948 at Sillamaee, in North-East Estonia, ca 190 km from Tallinn. Target product was uranium, mostly in form of yellow cake (U 3 O 8 ) for Soviet nuclear program. Uranium ore processing continued from 1948 to 1977, totally 4,013,000 tons of uranium ore were processed at Sillamaee plant. In early 1970s the plant introduced a new production line - rare earth elements. Rare earths were until 1991 produced from loparite (later from semi-processed loparite) - rare earths, niobium, tantalum and NORM-containing ore for Kola peninsula, Russia; later. All wastes were, as typical to hydrometallurgical processing all over the world, discharged to a large, 40 ha liquid waste depository - tailings pond, what in Sillamaee case was designed to discharge all liquid constituents slowly to the Baltic Sea. All uranium related activities were stopped in 1990, when only rare earth and rare metal production lines remained operational. The plant was 100 % privatized in 1997 and is today operated by Silmet Ltd., processing annually up to 8 000 tons of rare earth and 2000 tons of niobium and tantalum ores. Like all industries, inherited from Soviet times, Silmet plant is today facing a serious challenge to upgrading technologies towards waste minimizations process efficiency. The historical tailings pond, containing ca 1800 tons of natural uranium and ca 800 tons of thorium, was found geotechnically unstable and leaking to the Baltic Sea, in mid 90s. Being a problem of common Baltic concern, an international remediation project was initiated by Estonian Government and plant operator in 1998. In cooperation with Estonian, Finnish, Swedish, Danish and Norwegian Governments and with assistance by the European Union, the tailings pond will be environmentally remediated - dams stabilized and surface covered, by end of 2006. Close-down and environmental remediation of the tailings pond provides plant an ultimate challenge of

Alternative containers for low-level wastes containing large amounts of tritium

International Nuclear Information System (INIS)

Gause, E.P.; Lee, B.S.; MacKenzie, D.R.; Wiswall, R. Jr.

1984-11-01

High-activity tritiated waste generated in the United States is mainly composed of tritium gas and tritium-contaminated organic solvents sorbed onto Speedi-Dri which are packaged in small glass bulbs. Low-activity waste consists of solidified and adsorbed liquids. In this report, current packages for high-activity gaseous and low-activity adsorbed liquid wastes are emphasized with regard to containment potential. Containers for low-level radioactive waste containing large amounts of tritium need to be developed. An integrity may be threatened by: physical degradation due to soil corrosion, gas pressure build-up (due to radiolysis and/or biodegradation), rapid permeation of tritium through the container, and corrosion from container contents. Literature available on these points is summarized in this report. 136 references, 20 figures, 40 tables

Application of biosorbents in treatment of the radioactive liquid waste

International Nuclear Information System (INIS)

Ferreira, Rafael Vicente de Padua

2014-01-01

Radioactive liquid waste containing organic compounds need special attention, because the treatment processes available are expensive and difficult to manage. The biosorption is a potential treatment technique that has been studied in simulated wastes. The biosorption term is used to describe the removal of metals, non-metals and/or radionuclides by a material from a biological source, regardless of its metabolic activity. Among the potential biomasses, agricultural residues have very attractive features, as they allow for the removal of radionuclides present in the waste using a low cost biosorbent. The aim of this study was to evaluate the potential use of different biomass originating from agricultural products (coconut fiber, coffee husk and rice husk) in the treatment of real radioactive liquid organic waste. Experiments with these biomass were made including 1) Preparation, activation and characterization of biomasses; 2) Conducting biosorption assays; and 3) Evaluation of the product of immobilization of biomasses in cement. The biomasses were tested in raw and activated forms. The activation was carried out with diluted HNO 3 and NaOH solutions. Biosorption assays were performed in polyethylene bottles, in which were added 10 mL of radioactive waste or waste dilutions in deionized water with the same pH and 2% of the biomass (w/v). At the end of the experiment, the biomass was separated by filtration and the remaining concentration of radioisotopes in the filtrate was determined by ICP-OES and gamma spectrometry. The studied waste contains natural uranium, americium-241 and cesium-137. The adopted contact times were 30 min, 1, 2 and 4 hours and the concentrations tested ranged between 10% and 100%. The results were evaluated by maximum experimental sorption capacity and isotherm and kinetics ternary models. The highest sorption capacity was observed with raw coffee husk, with approximate values of 2 mg/g of U (total), 40 x 10 -6 mg/g of Am-241 and 50 x10 -9

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.)

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.)

Biosorption of uranium by Azolla, SP, Brazil

International Nuclear Information System (INIS)

Vieira, Ludmila C.; Alves, Eliakim G.; Marumo, Julio T.; Ferreira, Rafael V. de P.; Canevesi, Rafael L.S.; Silva, Edson A.

2015-01-01

Radioactive liquid waste needs special attention and requires suitable treatment before deposition. Among the potential technologies under development for the treatment of liquid radioactive wastes the biosorption has been highlighted by being an efficient and low cost technique. Biosorption process involves the exchange of ions contained in the biomass matrix by others present in solution. There are many biomasses that could be applied in treatment of radioactive wastes, for example, agricultural residues and macrophyte. The aim of this study is evaluate the ability of the Azolla sp., a floating aquatic plant, to absorb uranium in solution. Azolla sp. is a macrophyte that has been used to treat effluents containing heavy metals. The biosorption capacity of uranium by Azolla sp. was experimentally determined and modeled by isotherms. Experiments were performed to determine metal uptake, and then the solutions were analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES). The isotherms applied to model the data was Langmuir, Freundlich, Sips Toth, Redlich Peternson, Two-Site-Langmuir, Radke Prausnitz to develop a technique for the treatment of radioactive liquid waste generated at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), Brazil. (author)

Biosorption of uranium by Azolla, SP, Brazil

Energy Technology Data Exchange (ETDEWEB)

Vieira, Ludmila C.; Alves, Eliakim G.; Marumo, Julio T., E-mail: lcvieira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Ferreira, Rafael V. de P., E-mail: rafael@itatijuca.com [Itatijuca Biotech, Sao Paulo, SP (Brazil); Canevesi, Rafael L.S.; Silva, Edson A., E-mail: edson.silva2@unioeste.br [Universidade Estadual do Oeste Parana (UNIOESTE), Toledo, PR (Brazil)

2015-07-01

Radioactive liquid waste needs special attention and requires suitable treatment before deposition. Among the potential technologies under development for the treatment of liquid radioactive wastes the biosorption has been highlighted by being an efficient and low cost technique. Biosorption process involves the exchange of ions contained in the biomass matrix by others present in solution. There are many biomasses that could be applied in treatment of radioactive wastes, for example, agricultural residues and macrophyte. The aim of this study is evaluate the ability of the Azolla sp., a floating aquatic plant, to absorb uranium in solution. Azolla sp. is a macrophyte that has been used to treat effluents containing heavy metals. The biosorption capacity of uranium by Azolla sp. was experimentally determined and modeled by isotherms. Experiments were performed to determine metal uptake, and then the solutions were analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES). The isotherms applied to model the data was Langmuir, Freundlich, Sips Toth, Redlich Peternson, Two-Site-Langmuir, Radke Prausnitz to develop a technique for the treatment of radioactive liquid waste generated at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), Brazil. (author)

Processing method for liquid waste containing various kinds of radioactive material

International Nuclear Information System (INIS)

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

1996-01-01

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

Cost study on waste management at three model Canadian uranium mines

International Nuclear Information System (INIS)

1984-03-01

A waste management cost study was initiated to determine the capital and operating costs of three different uranium waste management systems which incorporate current technologies being used in Canadian uranium mining operations. Cost estimates were to be done to a thirty percent level of accuracy and were to include all waste management related costs of a uranium ore processing facility. Each model is based on an annual uranium production of 1,923,000 kg U (5,000,000 lbs U 3 O 8 ) with a total operating life of 20 years for the facility. The three models, A, B, and C, are based on three different uranium ore grades, 0.10 percent U 3 O 8 , 0.475 percent U 3 O 8 and 1.5 percent U 3 O 8 respectively. Yellowcake production is assumed to start in January 1984. Model A is based on a conceptual 7,180 tonne per day uranium ore processing facility and waste management system typical of uranium operations in the Elliot Lake area of northern Ontario with an established infrastructure. Model B is a 1.512 tonne per day operation based on a remote uranium operation typical of the Athabasca Basin properties in northern Saskatchewan. Model C is a 466 tonne per day operation processing a high-grade uranium ore containing arsenic and heavy metal concentrations typical of some northern Saskatchewan deposits

Radioactive liquid waste processing system

International Nuclear Information System (INIS)

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

1997-01-01

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

Method of cement-solidification of radioactive liquid wastes containing surfactant

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Yusa, H

1979-04-10

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

Evaluation of transport properties of nanofiltration membranes exposed to radioactive liquid waste

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Bastos, Edna T.R., E-mail: eemo@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeira, RJ (Brazil); Afonso, Julio C., E-mail: Julio@iq.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Inst. de Quimica. Dept. de Quimica Analitica

2011-07-01

The application of membrane separation processes (PSM) for treatment of radioactive waste requires the selection of a suitable membrane for the treatment of waste, as the membrane will be directly exposed to the radioactive liquid waste, and also exposed to ionizing radiation. The nanofiltration membrane is most suitable for treatment of radioactive waste, since it has high rejection of multivalent ions. Usually the membranes are made of polymers and depending on the composition of the waste, type and dose of radiation absorbed may be changes in the structure of the membrane, resulting in loss of its transport properties. We tested two commercial nanofiltration membranes: NF and SW Dow/Filmtec. The waste liquid used was obtained in the process of conversion of uranium hexafluoride gas to solid uranium dioxide, known as 'carbonated water'. The membranes were characterized as their transport properties (hydraulic permeability, permeate flux and salt rejection) before and after their immersion in the waste for 24 hours. The surface of the membranes was also evaluated by SEM and FTIR. It was observed that in both the porosity of the membrane selective layer was altered, but not the membrane surface charge, which is responsible for the selectivity of the membrane. The NF membranes and SW showed uranium ion rejection of 64% and 55% respectively. (author)

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

International Nuclear Information System (INIS)

Christensen, B. D.

1999-01-01

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

SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

International Nuclear Information System (INIS)

Rathbun, R.

1993-01-01

Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

Recovery of uranium and accompanying metals from various types of industrial wastes

International Nuclear Information System (INIS)

Chajduk, E.; Danko, B.; Gajda, D.; Zakrzewska, G.; Harasimowicz, M.; Bieluszka, P.

2014-01-01

On January 28"t"h 2014 the Program of Polish Nuclear Energy was signed by Polish Government. According to this program Poland has to secure a constant supply of uranium for Polish NPPs in the future. Uranium in Poland occurs in Vistula Spit area in sandstone rocks and Podlasie Depression area in black dictyonema shales, which are low grade ores. Scarce uranium resources stimulate interest in its recovery from secondary resources as potential raw materials. Industrial wastes and by-products were considered as a source of uranium in this studies. Apart from uranium other valuable metals (e.g. vanadium, molybdenum or lanthanides) were recovered to improve the economy of the process. Three types of industrial wastes were examined: flotation tailings from the copper industry, phosphoric acid from the fertilizer industry and fracturing fluid from shale gas exploitation. Metals from flotation tailings were separated in two steps: 1) acidic leaching of the flotation waste using sulfuric acid solution and 2) separation of metals by ion-exchange chromatography. All the liquid samples were analyzed by ICP-MS method to determine the separation efficiency of the process. Uranium was recovered from phosphoric acid by high-pressure membrane filtration or by extraction/stripping integrated processes applying membrane modules Liquid-Cel® Extra-Flow (Celgard). Aqueous solutions after hydraulic fracturing are very diverse in terms of chemical composition, depending on borehole and fracturing technology applied. The content of various substances in backflow fluid depends on mechanical behavior and chemical composition of shale. Organic matter content in this type of waste did not exceed 1% usually, but the salinity is high. Initially, organic pollutants were removed and next the fluid was purified by combined various ion-exchangers. Individual metals were selectively eluted from ion-exchanger by combination of different eluents. The content of metals in samples was analyzed by ICP

Treatment of low level radioactive liquid waste containing appreciable concentration of TBP degraded products.

Science.gov (United States)

Valsala, T P; Sonavane, M S; Kore, S G; Sonar, N L; De, Vaishali; Raghavendra, Y; Chattopadyaya, S; Dani, U; Kulkarni, Y; Changrani, R D

2011-11-30

The acidic and alkaline low level radioactive liquid waste (LLW) generated during the concentration of high level radioactive liquid waste (HLW) prior to vitrification and ion exchange treatment of intermediate level radioactive liquid waste (ILW), respectively are decontaminated by chemical co-precipitation before discharge to the environment. LLW stream generated from the ion exchange treatment of ILW contained high concentrations of carbonates, tributyl phosphate (TBP) degraded products and problematic radio nuclides like (106)Ru and (99)Tc. Presence of TBP degraded products was interfering with the co-precipitation process. In view of this a modified chemical treatment scheme was formulated for the treatment of this waste stream. By mixing the acidic LLW and alkaline LLW, the carbonates in the alkaline LLW were destroyed and the TBP degraded products got separated as a layer at the top of the vessel. By making use of the modified co-precipitation process the effluent stream (1-2 μCi/L) became dischargeable to the environment after appropriate dilution. Based on the lab scale studies about 250 m(3) of LLW was treated in the plant. The higher activity of the TBP degraded products separated was due to short lived (90)Y isotope. The cement waste product prepared using the TBP degraded product was having good chemical durability and compressive strength. Copyright © 2011 Elsevier B.V. All rights reserved.

Bacterial leaching of waste uranium materials.

Science.gov (United States)

Barbic, F F; Bracilović, D M; Krajincanić, B V; Lucić, J L

1976-01-01

The effect of ferrobacteria and thiobacteria on the leaching of waste uranium materials from which 70-80% of uranium was previously leached by classical chemical hydrometallurgical procedure has been investigated. The bacteria used are found in the ore and the mine water of Zletovska River locality, Yugoslavia. Parameters of biological leaching were examined in the laboratory. Leaching conditions were changed with the aim of increasing the amount of uranium leached. The effect of pyrite added to the waste materials before the beginning of leaching has also been examined. Uranium leaching is directly proportional to the composition and number of ferrobacteria and thiobacteria, and increased by almost twice the value obtained from the same starting materials without using bacteria. Increased sulphuric acid concentrations stimulate considerably the rate of leaching. Uranium leaching is increased up to 20% while sulphuric acid consumption is simultaneously decreased by the addition of pyrite. Uranium concentrations in starting waste materials used for leaching were extremely low (0.0278 and 0.372% U) but about 60% recovery of uranium was obtained, with relatively low consumption of sulphuric acid.

Bacterial leaching of waste uranium materials

International Nuclear Information System (INIS)

Barbic, F.F.; Bracilovic, D.M.; Krajincanic, B.V.; Lucic, J.L.

1976-01-01

The effect of ferrobacteria and thiobacteria on the leaching of waste uranium materials from which 70-80% of uranium was previously leached by classical chemical hydrometallurgical procedure has been investigated. The bacteria used are found in the ore and the mine water of Zletovska River locality, Yugoslavia. Parameters of biological leaching were examined in the laboratory. Leaching conditions were changed with the aim of increasing the amount of uranium leached. The effect of pyrite added to the waste materials before the beginning of leaching has also been examined. Uranium leaching is directly proportional to the composition and number of ferrobacteria and thiobacteria, and increased by almost twice the value obtained from the same starting materials without using bacteria. Increased sulphuric acid concentrations stimulate considerably the rate of leaching. Uranium leaching is increased up to 20% while sulphuric acid consumption is simultaneously decreased by the addition of pyrite. Uranium concentrations in starting waste materials used for leaching were extremely low (0.0278 and 0.0372% U) but about 60% recovery of uranium was obtained, with relatively low consumption of sulphuric acid. (author)

Decontamination flowsheet development for a waste oil containing mixed radioactive contaminants

International Nuclear Information System (INIS)

Vijayan, S.; Buckley, L.P.

1993-01-01

The majority of waste oils contaminated with both radioactive and hazardous components are generated in nuclear power plant, research lab. and uranium-refinery operations. The waste oils are complex, requiring a detailed examination of the waste management strategies and technology options. It may appear that incineration offers a total solution, but this may not be true in all cases. An alternative approach is to decontaminate the waste oils to very low contaminant levels, so that the treated oils can be reused, burned as fuel in boilers, or disposed of by commercial incineration. This paper presents selected experimental data and evaluation results gathered during the development of a decontamination flowsheet for a specific waste oil stores at Chalk River Labs. (CRL). The waste oil contains varying amounts of lube oils, grease, paint, water, particulates, sludge, light chloro- and fluoro-solvents, polychlorinated biphenyls (PCB), complexing chemicals, uranium, chromium, iron, arsenic and manganese. To achieve safe management of this radioactive and hazardous waste, several treatment and disposal methods were screened. Key experiments were performed at the laboratory-scale to confirm and select the most appropriate waste-management scheme based on technical, environmental and economic criteria. The waste-oil-decontamination flowsheet uses a combination of unit operations, including prefiltration, acid scrubbing, and aqueous-leachage treatment by precipitation, microfiltration, filter pressing and carbon adsorption. The decontaminated oil containing open-quotes de minimisclose quotes levels of contaminants will undergo chemical destruction of PCBs and final disposal by incineration. The recovered uranium will be recycled to a uranium milling process

Joule-Heated Ceramic-Lined Melter to Vitrify Liquid Radioactive Wastes Containing Am241 Generated From MOX Fuel Fabrication in Russia

International Nuclear Information System (INIS)

Smith, E C; Bowan II, B W; Pegg, I; Jardine, L J

2004-01-01

The governments of the United Stated of America and the Russian Federation (RF) signed an Agreement September 1, 2000 to dispose of weapons plutonium that has been designated as no longer required for defense purposes. The Agreement declares that each country will disposition 34MT of excess weapons grade plutonium from their stockpiles. The preferred disposition technology is the fabrication of mixed oxide (MOx) fuel for use or burning in pressurized water reactors to destroy the plutonium. Implementation of this Agreement will require the conversion of plutonium metal to oxide and the fabrication of MOx fuel within the Russian Federation. The MOx fuel fabrication and metal to oxide conversion processes will generate solid and liquid radioactive wastes containing trace amounts of plutonium, neptunium, americium, and uranium requiring treatment, storage, and disposal. Unique to the Russian MOx fuel fabrication facility's flow-sheet is a liquid waste stream with high concentrations (∼1 g/l) of 241 Am and non radioactive silver. The silver is used to dissolve PuO 2 feed materials to the MOx fabrication facility. Technical solutions are needed to treat and solidify this liquid waste stream. Alternative treatment technologies for this liquid waste stream are being evaluated by a Russian engineering team. The technologies being evaluated include borosilicate and phosphate vitrification alternatives. The evaluations are being performed at a conceptual design level of detail under a Lawrence Livermore National Laboratory (LLNL) contract with the Russian organization TVEL using DOE NA-26 funding. As part of this contract, the RF team is evaluating the technical and economic feasibility of the US borosilicate glass vitrification technology based on a Duratek melter to solidify this waste stream into a form acceptable for storage and geologic disposal. The composition of the glass formed from treating the waste is dictated by the concentration of silver and americium it

Treatment of low alpha activity liquid wastes

International Nuclear Information System (INIS)

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

1984-01-01

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

Radioactive liquid waste solidifying device

International Nuclear Information System (INIS)

Uchiyama, Yoshio.

1987-01-01

Purpose: To eliminate the requirement for discharge gas processing and avoid powder clogging in a facility suitable to the volume-reducing solidification of regenerated liquid wastes containing sodium sulfate. Constitution: Liquid wastes supplied to a liquid waste preheater are heated under a pressure higher than the atmospheric pressure at a level below the saturation temperature for that pressure. The heated liquid wastes are sprayed from a spray nozzle from the inside of an evaporator into the super-heated state and subjected to flash distillation. They are further heated to deposit and solidify the solidification components in the solidifying evaporation steams. The solidified powder is fallen downwardly and heated for removing water content. The recovered powder is vibrated so as not to be solidified and then reclaimed in a solidification storage vessel. Steams after flash distillation are separated into gas, liquid and solids by buffles. (Horiuchi, T.)

Application of liquid-liquid extraction in uranium hydrometallurgy (Paper No. : V-1)

Energy Technology Data Exchange (ETDEWEB)

Murthy, T K.S.; Koppiker, K S

1979-01-01

Uranium recovery from the ores is carried out exclusively by hydrometallurgical techniques. The initial solubilisation of uranium is achieved by either sodium carbonate or sulphuric acid leaching, the latter being more common. Further purification and upgrading of uranium from the sulphate liquors is carried out by an ion-exchange process. Solid resin type anion exchangers or liquid ion-exchangers are employed. The processing of uranium liquors is, perhaps, the first major application of liquid-liquid extraction in metal recovery. Organophosphoric acids were initially used but later the long-chain aliphatic amines have superseded them. The amine extraction system has been widely studied and several variations are now known. Chloride, nitrate, carbonate or sulphate or acid stripping can be used for getting back the uranium into the aqueous phase. Combination of ion exchange (resin type) and solvent extraction processes called Eluex processes are developed for special applications. Studies have also been made of solvent extraction of uranium from leach pulps instead of clear liquors. Tributylphosphate has found wide application in the refining of uranium concentrates to meet the stringent needs of nuclear purity. liquid-liquid extraction is, perhaps, the only successful technique for the recovery of uranium, as by-product, from wet-process phosphoric acid. This has opened up a new source of uranium.

Catalytic properties of mineral ion-exchangers previously used in uranium ore wastes treatment. I. Ammoniac octylation in gaseous phase using Y-type faujasite catalysts containing mainly uranium

International Nuclear Information System (INIS)

Azzouz, A.; Nibou, D.; Abbad, B.; Achache, M.

1990-06-01

Y-type faujasite, previously used in purifying aqueous wastes containing radioactive elements, are studied as catalysts in heterogeneous reactions such as octanol amination. This process consists in alkylating ammoniac NH 3 by octanol-1 in a flow reactor containing a fixed bed of pelleted catalysts. Acidic catalysts like Y-type faujasite impregnated with low concentration leaching solution containing between 10 and 1000 ppm of uranium show interesting activities and selectivities in yielding primary, secondary and tertiary amines. High octanol conversions were obtained reaching 60 mol.% for appreciable selectivities for producing amines of about 60-80 mol.%. Compared to the non-active fresh zeolite, the catalysts obtained present good activities essentially due to the presence of uranyl cations UO 2 ++ . The latter seems to play the main role in activating the zeolite by enhancing its Bronsted surface acidity. Another interest of this process consists in obtaining Trioctylamine (TOA), an important extraction agent in uranium hydrometallurgy or its derivatives (MOA and DOA) which are very used in chemical industry

Bioremediation of uranium contaminated soils and wastes

International Nuclear Information System (INIS)

Francis, A.J.

1998-01-01

Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (1) stabilization of uranium and toxic metals with reduction in waste volume and (2) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs

Application of macrophytes as biosorbents for radioactive liquid waste treatment; Aplicacao de macrofitas como biossorventes no tratamento de rejeitos radioativos liquidos

Energy Technology Data Exchange (ETDEWEB)

Vieira, Ludmila Cabreira

2016-07-01

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

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

International Nuclear Information System (INIS)

Zainus Salimin; Endang Nuraeni; Mirawaty; Tarigan, Cerdas

2010-01-01

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

Radioactive waste management

International Nuclear Information System (INIS)

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

1975-08-01

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

Uranium oxide recovering method

International Nuclear Information System (INIS)

Ota, Kazuaki; Takazawa, Hiroshi; Teramae, Naoki; Onoue, Takeshi.

1997-01-01

Nitrates containing uranium nitrate are charged in a molten salt electrolytic vessel, and a heat treatment is applied to prepare molten salts. An anode and a cathode each made of a graphite rod are disposed in the molten salts. AC voltage is applied between the anode and the cathode to conduct electrolysis of the molten salts. Uranium oxides are deposited as a recovered product of uranium, on the surface of the anode. The nitrates containing uranium nitrate are preferably a mixture of one or more nitrates selected from sodium nitrate, potassium nitrate, calcium nitrate and magnesium nitrate with uranium nitrate. The nitrates may be liquid wastes of nitrates. The temperature for the electrolysis of the molten salts is preferably from 150 to 300degC. The voltage for the electrolysis of the molten salts is preferably an AC voltage of from 2 to 6V, more preferably from 4 to 6V. (I.N.)

Biochemistry Oxidation Process for Treatment the Simulation of Organic Liquid Radioactive Waste

International Nuclear Information System (INIS)

Gunandjar; Zainus Salimin; Sugeng Purnomo; Ratiko

2010-01-01

The nuclear industry activities generate the organic liquid wastes such as detergent waste from laundry, solvent waste of 30% TBP (tri-n-butyl phosphate) in kerosene from purification or recovery of uranium from rejection of nuclear fuel element fabrication, and solvent waste containing D 2 EHPA (di-2-ethyl hexyl phosphoric acid) and TOPO (trioctyl phospine oxide) in kerosene from phosphoric acid purification. The wastes are included in category of the hazard and poison materials which also radioactive, so that the wastes have to be treated to detoxification of the hazard and poison materials and decontamination of the radionuclides. The research of biochemistry oxidation process for treatment the simulation of organic liquid radioactive waste from laundry using mixture of aerobe bacteria of bacillus sp, pseudomonas sp, arthrobacter sp, and aeromonas sp have been carried out. The waste containing detergent 1,496 g/Litre, activity 10 -1 Ci/m 3 , with COD (Chemical Oxygen Demand) 128, BOD (Biological Oxygen Demand) 68 and TSS (Total Suspended Solid) 1000 ppm, it is treated by biochemistry oxidation with addition of bacteria which be fed nutrition of nitrogen and phosphor, and aeration. The result show that the bacteria can decompose the detergent to become carbon dioxyde and water so that can fulfill the quality standard of water group-B with content of BOD and COD are 6 and 10 ppm respectively, the time of decomposition is needed 106 hours to be fulfill the quality standard of water. The longer of process time will give bigger the total solid content in sludge, because the biomass generated from the colony of bacteria which life and dead to so much. (author)

About possibility of uranium industry wastes reprocessing in Tajikistan

International Nuclear Information System (INIS)

Khakimov, N.; Barotov, B.B.; Mirsaidov, I.U.; Barotov, A.M; Akhmedov, M.Z.

2012-01-01

One of the main basic fields of economy in Tajikistan is mining industry. Its development in the past lead to accumulation of huge amount of wastes basically from uranium reprocessing enterprises, containing radionuclides in anthropogenic highly concentrations (basically uranium-thorium chain) and other harmful substances. They are located in zones very close to residential as well as in upper course of water inflow of such main rivers of the region as Amudarya and Syrdarya. Sulphates, heavy metals, cyanides and others (basically with water flow) released to the environment besides uranium reprocessing wastes and other mining enterprises. This makes it necessary to restore in the region the complex coordinated monitoring programs with the purpose of their actual assessment and potential impact on environment as well as priority justification of possible remediation measures. One of the important are balance assessment and trans-boundary radionuclide re-deposition of uranium-thorium chain and other toxic elements in Syrdarya and Amudarya rivers with the purpose of regional formation character revealing of radiation and other ecological risks for saving the normal vital activity in the region.

Study on the treatment of waste waster containing uranium by organic modified vermiculite

International Nuclear Information System (INIS)

Liu Wenjuan; Zeng Yanhong

2012-01-01

The adsorption capability of uranium on organic modified Vermiculite was studied. The influence factors of the amount of adsorbent, initial pH, initial concentration of uranium and adsorption time have been investigated too. Through the orthogonal test, the primary factors of impacting the adsorption treatment can be obtained. Finally, the preliminary research and analysis on the principle adsorption of organic modified vermiculite test of uranium have been conducted. The results show that: Modifying Vermiculite by CTMAB makes Vermiculite adsorption capacity stronger when treating solution containing uranium. Combined flocculants with vermiculite to treat with low concentration of uranium solution has synergy, significantly enhancing its adsorption capacity. The impact factors of organic modified vermiculite's adsorption of uranium are adsorbent dosage, pH, initial concentration of uranium solution and adsorption time. The best adsorption pH is between 5∼6.5. (authors)

NEW CRITERIA FOR ASSIGNING WASTE CONTAINING TECH-NOGENIC RADIONUCLIDES TO THE RADIOACTIVE WASTE

Directory of Open Access Journals (Sweden)

I. K. Romanovich

2010-01-01

Full Text Available The article contains detailed description of criteria for assigning of liquid and gaseous industrial waste containing technogenicradionuclides to the radioactive waste, presented in the new Basic Sanitary Rulesof Radiation Safety (OSPORB-99/2010. The analysisof shortcomings and discrepancies of the previously used in Russia system of criteria for assigning waste to the radioactive waste is given.

Subsides for optimization of transfer of radioactive liquid waste from 99MO production plant to the waste treatment facility

International Nuclear Information System (INIS)

Rego, Maria Eugenia de Melo; Vicente, Roberto; Hiromoto, Goro

2013-01-01

The increasing need for radioisotopes lead Brazil to consider the domestic production of 99 Mo from fission of low enriched uranium targets. In order to meet the present demand of 99m Tc generators the planned 'end of irradiation' activity of 99 Mo is about 170 TBq per week. The radioactive waste from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the waste were predicted based on the fission yield and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production process and schedule, already established by the project management. The transfer of the waste from the production plant to the treatment facility will be done by means of special shielded packages. In the present study, the commercially available code Scale 6.0 was used to simulate the irradiation of the targets and the decay of radioactive products, assuming that an alkaline dissolution process would be performed on the targets before the removal and purification of 99 Mo. The assessment of the shielding required for the packages containing liquid waste was done using MicroShield 9 code. The results presented here are part of a project that aims at contributing to the design of the waste management system for the 99 Mo production facility. (author)

Titanate ceramics for immobilisation of uranium-rich radioactive wastes arising from {sup 99}Mo production

Energy Technology Data Exchange (ETDEWEB)

Carter, M.L.; Li, H. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia); Zhang, Y. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)], E-mail: yzx@ansto.gov.au; Vance, E.R.; Mitchell, D.R.G. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)

2009-02-28

Uranium-rich liquid wastes arising from UO{sub 2} targets which have been neutron-irradiated to generate medical radioisotopes such as {sup 99m}Tc require immobilisation. A pyrochlore-rich hot isostatically pressed titanate ceramic can accommodate at least 40 wt% of such waste expressed on an oxide basis. In this paper, the baseline waste form composition (containing 40 wt% UO{sub 2}) was adjusted in two ways: (a) varying the UO{sub 2} loading with constant precursor oxide materials, (b) varying the precursor composition with constant waste loading of UO{sub 2}. This resulted in the samples having a similar phase assemblage but the amounts of each phase varied. The oxidation states of U in selected samples were determined using diffuse reflection spectroscopy (DRS) and electron energy loss spectroscopy (EELS). Leaching studies showed that there was no significant difference in the normalised elemental release rates and the normalised release rates are comparable with those from synroc-C. This demonstrates that waste forms based on titanate ceramics are robust and flexible for the immobilisation of U-rich waste streams from radioisotope processing.

Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

International Nuclear Information System (INIS)

Forsberg, C.W.

1996-01-01

A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

PNGMDR 2013-2015. The management of liquid and gaseous tritium-containing wastes from the non-electronuclear sector. Progress status by the end of 2013

International Nuclear Information System (INIS)

2014-01-01

After having briefly evoked the characteristics of tritium-containing radioactive wastes, and the associated issues for storage or warehousing, and also outlined that warehousing solutions are designed for solid wastes whereas tritium is often used under its liquid and gaseous form, this report addresses the case of these liquid and gaseous tritium-containing radioactive wastes. A first part addresses gaseous tritium-containing wastes, discusses their inventory without taking tritium-containing objects belonging to the National Defence. Thus, Tritium is present in lightning arresters, in radio-luminescent objects, in emergency exit panels. The present location of these wastes is commented, and the constraints related to their taking into charge by the Aube waste centre are discussed: issue of releases, regulatory requirements, acceptance technical specifications, determination of the gaseous Tritium LAS (limit of acceptance of a sealed source). The report proposes an overview of alternative pathways to storage in the Aube storage centre: destruction and discharge when authorized, use of new equipment developed by the ANDRA, taking over by the country of origin or recycling, decay warehousing. The next part of the report addresses liquid tritium-containing wastes, proposes a brief inventory, and briefly evokes two options: combustion, or discharge when authorized

Radiolytic degradation of octachlorodibenzo-p-dioxin and octachlorodibenzofuran in organic solvents and treatment of dioxin-containing liquid wastes

International Nuclear Information System (INIS)

Zhao Changli; Hirota, Koichi; Taguchi, Mitsumasa; Takigami, Machiko; Kojima, Takuji

2007-01-01

Degradations of octachlorodibenzo-p-dioxin (OCDD) and octachlorodibenzofuran (OCDF) were studied by 60 Co γ-ray in organic solvents: ethanol, n-nonane, and toluene. Both OCDD and OCDF were degraded more efficiently in ethanol than in n-nonane or toluene. The degradation is mainly attributed to electrons and in part to solvent radicals. The addition of ethanol to dioxin-containing liquid wastes enhanced effectively the degradation of dioxins; the liquid wastes did not exhibit the dioxin toxicity at a dose of 100 kGy

Treatment of radioactive organics liquid wastes

International Nuclear Information System (INIS)

Morales Galarce, Tania

1999-01-01

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

Processing method for radioactive liquid waste

International Nuclear Information System (INIS)

Yasumura, Keijiro

1991-01-01

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

Subsides for optimization of transfer of radioactive liquid waste from {sup 99}MO production plant to the waste treatment facility

Energy Technology Data Exchange (ETDEWEB)

Rego, Maria Eugenia de Melo; Vicente, Roberto; Hiromoto, Goro, E-mail: maria.eugenia@ipen.br, E-mail: rvicente@ipen.br, E-mail: hiromoto@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The increasing need for radioisotopes lead Brazil to consider the domestic production of {sup 99}Mo from fission of low enriched uranium targets. In order to meet the present demand of {sup 99m}Tc generators the planned 'end of irradiation' activity of {sup 99}Mo is about 170 TBq per week. The radioactive waste from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the waste were predicted based on the fission yield and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production process and schedule, already established by the project management. The transfer of the waste from the production plant to the treatment facility will be done by means of special shielded packages. In the present study, the commercially available code Scale 6.0 was used to simulate the irradiation of the targets and the decay of radioactive products, assuming that an alkaline dissolution process would be performed on the targets before the removal and purification of {sup 99}Mo. The assessment of the shielding required for the packages containing liquid waste was done using MicroShield 9 code. The results presented here are part of a project that aims at contributing to the design of the waste management system for the {sup 99}Mo production facility. (author)

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

International Nuclear Information System (INIS)

1995-01-01

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

Optimization and validation of a chemical process for uranium, mercury and cesium leaching from cemented radioactive wastes

International Nuclear Information System (INIS)

Reynier, N.; Riveros, P.; Lastra, R.; Laviolette, C.; Bouzoubaa, N.; Chapman, M.

2015-01-01

Atomic Energy of Canada Limited (AECL) is developing a treatment and long-term management strategy for a legacy cemented radioactive waste that contains uranium, mercury and fission products. Extracting the uranium would be advantageous for decreasing the waste classification and reducing the cost of long-term management. Consequently, there are safety and economic and environmental incentives for the extraction of uranium, mercury and cesium before subjecting the cemented waste to a stabilization process. The mineralogical analysis of the surrogate cemented waste (SCW) indicated that uranium forms calcium uranate, CaUO 4 , occurring as layers of several millimeters or as grains of 20 μm. Hg is found mostly as large (∼50 μm) and small grains (5-8 μm) of HgO. The chemical leachability of three key elements (U, Hg, and Cs) from a SCW was studied with several leaching materials. The results showed that the most promising approach to leach and recover U, Hg, and Cs is the direct leaching of the SCW with H 2 SO 4 in strong saline media. Operating parameters such as particle size, temperature, pulp density, leaching time, acid and salt concentrations, number of leaching/rinsing step, etc. were optimized to improve key elements solubilization. Sulfuric leaching in saline media of a SCW (U5) containing 1182 ppm of U, 1598 ppm of Hg, and 7.9 ppm of Cs in the optimized conditions allows key elements recovery of 98.5 ± 0.4%, 96.6 ± 0.1%, and 93.8 ± 1.1% of U, Hg, and Cs respectively. This solubilization process was then applied in triplicate to seven other SCW prepared with different cement, liquid ratio and at different aging time and temperature. Concentrated sulfuric acid is added to the slurry until the pH is about 2, which causes the complete degradation of cement and the formation of CaSO 4 . At this pH, the acid consumption is moderate and the formation of amorphous silica gel is avoided. Sulfuric acid is particularly useful because it produces a leachate that

Assessment of nonpoint source chemical loading potential to watersheds containing uranium waste dumps and human health hazards associated with uranium exploration and mining, Red, White, and Fry Canyons, southeastern Utah, 2007

Science.gov (United States)

Beisner, Kimberly R.; Marston, Thomas M.; Naftz, David L.; Snyder, Terry; Freeman, Michael L.

2010-01-01

During May, June, and July 2007, 58 solid-phase samples were collected from abandoned uranium mine waste dumps, background sites, and adjacent streambeds in Red, White, and Fry Canyons in southeastern Utah. The objectives of this sampling program were to (1) assess the nonpoint-source chemical loading potential to ephemeral and perennial drainage basins from uranium waste dumps and (2) assess potential effects on human health due to recreational activities on and around uranium waste dumps on Bureau of Land Management property. Uranium waste-dump samples were collected using solid-phase sampling protocols. After collection, solid-phase samples were homogenized and extracted in the laboratory using a leaching procedure. Filtered (0.45 micron) water samples were obtained from the field leaching procedure and were analyzed for major and trace elements at the Inductively Coupled Plasma-Mass Spectrometry Metals Analysis Laboratory at the University of Utah. A subset of the solid-phase samples also were digested with strong acids and analyzed for major ions and trace elements at the U.S. Geological Survey Geologic Division Laboratory in Denver, Colorado. For the initial ranking of chemical loading potential for uranium waste dumps, results of leachate analyses were compared with existing aquatic-life and drinking-water-quality standards. To assess potential effects on human health, solid-phase digestion values for uranium were compared to soil screening levels (SSL) computed using the computer model RESRAD 6.5 for a probable concentration of radium. One or more chemical constituents exceeded aquatic life and drinking-water-quality standards in approximately 64 percent (29/45) of the leachate samples extracted from uranium waste dumps. Most of the uranium waste dump sites with elevated trace-element concentrations in leachates were located in Red Canyon. Approximately 69 percent (31/45) of the strong acid digestible soil concentration values were greater than a calculated

Physico-chemical basics for production of uranium concentrate from wastes of hydrometallurgical plants and technical waters

International Nuclear Information System (INIS)

Mirsaidov, I.; Nazarov, K.

2014-01-01

Physico-chemical and technological basics for reprocessing of uranium industry wastes of Northern Tajikistan shows that the most perspective site for reprocessing is Chkalovkst tailings wastes. The engineering and geological conditions and content of radionuclides in wastes were investigated. It was determined that considered by radioactivity the wastes are low activity and they can be reprocessed for the purpose of U_3O_8 production. Characteristics of mine and technical waters of uranium industry wastes were studied. Characteristics of mine and technical waters of Kiik-Tal and Istiklol city (former Taboshar) showed the expediency of uranium oxide extraction from them. The reasons for non-additional recovery extraction from dumps of SE “Vostokredmet” by classical methods of uranium leaching are studied. The kinetics of sulfuric leaching of residues from anthropogenic deposit of Map 1-9 (Chkalovsk City) were also investigated. Further investigations are to reveal the flow mechanism process of sulfuric leaching of residues and to enable the selection of a radiation regime for U_3O_8 production. The kinetics of sorption process of uranium extraction from mine and technical waters of uranium industry wastes were studied. High sorption properties of apricot shell compared to other sorbents were revealed. A basic process flow diagram for reprocessing of uranium tailing wastes was developed as well as diagrams for uranium extraction from mine and technical waters from uranium industry wastes. The process consists of the following stages: acidification, sorption, burning, leaching, sedimentation, filtration and drying. The possibility of uranium extraction from natural uranic waters of a complicated salt composition was considered. Investigations revealed that uranium extraction from brines containing chloride ion is possible. A developed uranium extraction scheme from Sasik-Kul lake’s brine consists of the following main stages: evaporation, leaching, chloride

Optimization and validation of a chemical process for uranium, mercury and cesium leaching from cemented radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Reynier, N.; Lastra, R.; Laviolette, C.; Bouzoubaa, N., E-mail: nicolas.reynier@canada.ca [Natural Resources Canada, CanmetMINING, Ottawa, Ontario (Canada); Chapman, M. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2015-12-15

Canadian Nuclear Laboratories (CNL) is developing a treatment and long-term management strategy for a legacy cemented radioactive waste that contains uranium, mercury, and fission products. Extracting the uranium would be advantageous for decreasing the waste classification and reducing the cost of long-term management. The chemical leachability of 3 key elements (U, Hg, and Cs) from a surrogate cemented waste (SCW) was studied with several lixiviants. The results showed that the most promising approach to leach and recover U, Hg, and Cs is the direct leaching of the SCW with H{sub 2}SO{sub 4} in strong saline media. Operating parameters such as particle size, temperature, pulp density, leaching time, acid and salt concentrations, number of leaching/washing steps, etc. were optimized to improve key elements solubilization. Sulfuric leaching in saline media of a SCW (U5) containing 1182 ppm of U, 1598 ppm of Hg, and 7.9 ppm of Cs in the optimized conditions allows key elements solubilisation of 98.5 ± 0.4%, 96.6 ± 0.1%, and 93.8 ± 1.1% of U, Hg, and Cs, respectively. This solubilization process was then applied in triplicate to 7 other SCWs prepared with different cements, liquid ratios, and at different aging times and temperatures. Concentrated sulfuric acid is added to the slurry until the pH is about 2, which causes the complete degradation of cement and the formation of CaSO{sub 4}. Sulfuric acid is particularly useful because it produces a leachate that is amenable to conventional ion exchange technology for the separation and recovery of uranium. (author)

Filters for radioactive liquid wastes

International Nuclear Information System (INIS)

Koshiba, Yukihiko; Kawashima, Akio

1980-01-01

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

Process development for treatment of fluoride containing wastes

Energy Technology Data Exchange (ETDEWEB)

Singh, Mahesh; Kanvinde, V Y [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Many chemical and metallurgical industries generate liquid wastes containing high values of fluorides in association of nitrates and other metals. Due to harmful effects of fluorides these type of wastes can not be disposed off in the environment without proper treatment. Bench-scale laboratory experiments were conducted to develop a process scheme to fix the fluorides as non-leachable solid waste and fluoride free treated liquid waste for their disposal. To optimize the important parameters, simulated synthetic and actual wastes were used. For this study, three waste streams were collected from Nuclear Fuel Complex, Hyderabad. (author). 6 tabs., 1 fig.

XPS and ion beam scattering studies of leaching in simulated waste glass containing uranium

International Nuclear Information System (INIS)

Karim, D.P.; Pronko, P.P.; Marcuso, T.L.M.; Lam, D.J.; Paulikas, A.P.

1980-01-01

Glass samples (consisting of 2 mole % UO 3 dissolved in a number of complex borosilicate simulated waste glasses including Battelle 76-68) were leached for varying times in distilled water at 75 0 C. The glass surfaces were examined before and after leaching using x-ray photoemission spectroscopy and back-scattered ion beam profiling. Leached samples showed enhanced surface layer concentrations of several elements including uranium, titanium, zinc, iron and rare earths. An experiment involving the leaching of two glasses in the same vessel showed that the uranium surface enhancement is probably not due to redeposition from solution

Bioprecipitation of uranium from alkaline waste solutions using recombinant Deinococcus radiodurans

Energy Technology Data Exchange (ETDEWEB)

Kulkarni, Sayali; Ballal, Anand; Apte, Shree Kumar, E-mail: aptesk@barc.gov.in

2013-11-15

Highlights: • Deinococcus radiodurans was genetically engineered to overexpress alkaline phosphatase (PhoK). • Deino-PhoK bioprecipitated U efficiently over a wide range of input U concentration. • A maximal loading of 10.7 g U/g of biomass at 10 mM input U was observed. • Radioresistance and U precipitation by Deino-PhoK remained unaffected by γ radiation. • Immobilization of Deino-PhoK facilitated easy separation of precipitated U. -- Abstract: Bioremediation of uranium (U) from alkaline waste solutions remains inadequately explored. We engineered the phoK gene (encoding a novel alkaline phosphatase, PhoK) from Sphingomonas sp. for overexpression in the radioresistant bacterium Deinococcus radiodurans. The recombinant strain thus obtained (Deino-PhoK) exhibited remarkably high alkaline phosphatase activity as evidenced by zymographic and enzyme activity assays. Deino-PhoK cells could efficiently precipitate uranium over a wide range of input U concentrations. At low uranyl concentrations (1 mM), the strain precipitated >90% of uranium within 2 h while a high loading capacity of around 10.7 g U/g of dry weight of cells was achieved at 10 mM U concentration. Uranium bioprecipitation by Deino-PhoK cells was not affected in the presence of Cs and Sr, commonly present in intermediate and low level liquid radioactive waste, or after exposure to very high doses of ionizing radiation. Transmission electron micrographs revealed the extracellular nature of bioprecipitated U, while X-ray diffraction and fluorescence analysis identified the precipitated uranyl phosphate species as chernikovite. When immobilized into calcium alginate beads, Deino-PhoK cells efficiently removed uranium, which remained trapped in beads, thus accomplishing physical separation of precipitated uranyl phosphate from solutions. The data demonstrate superior ability of Deino-PhoK, over earlier reported strains, in removal of uranium from alkaline solutions and its potential use in

Sampling and characterization of radioactive liquid wastes

International Nuclear Information System (INIS)

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

2017-09-01

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

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

International Nuclear Information System (INIS)

Sherwood, D.J.

1994-03-01

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

Liquid-liquid extraction and separation studies of uranium(VI)

International Nuclear Information System (INIS)

Langade, A.D.; Shinde, V.M.

1980-01-01

Separation of uranium(VI) from iron(III), molybdenum(VI), vanadium(V), bismuth(III), zirconium(IV) and thorium(IV) is achieved by liquid-liquid extraction with 4-methyl-3-pentene-2-one (mesityl oxide; MeO) from sodium salicylate media (0.1M, pH 6.0). The extracted species is UO 2 (HO.C 6 H 4 COO) 2 .2MeO. A procedure for separating 50 μg of uranium from mg amounts of the other metals is described. (author)

The measurement test of uranium in a uranium-contaminated waste by passive gamma-rays measurement method

CERN Document Server

Sukegawa, Y; Ohki, K; Suzuki, S; Yoshida, M

2002-01-01

This report is completed about the measurement test and the proofreading of passive gamma - rays measurement method for Non - destructive assay of uranium in a uranium-contaminated waste. The following are the results of the test. 1) The estimation of the amount of uranium by ionization survey meter is difficult for low intensity of gamma-rays emitted from uranium under about 50g. 2) The estimation of the amount of uranium in the waste by NaI detector is possible in case of only uranium, but the estimation from mixed spectrums with transmission source (60-cobalt) is difficult to confirm target peaks. 3) If daughter nuclides of uranium and thorium chain of uranium ore exist, measurement by NaI detector is affected by gamma-rays from the daughter nuclides seriously-As a result, the estimation of the amount of uranium is difficult. 4) The measurement of uranium in a uranium-contaminated waste by germanium detector is possible to estimate of uranium and other nuclides. 5) As to estimation of the amount of uranium...

Uranium hexafluoride liquid thermal expansion, elusive eutectic with hydrogen fluoride, and very first production using chlorine trifluoride

Energy Technology Data Exchange (ETDEWEB)

Rutledge, G.P. [Central Environmental, Inc., Anchorage, AK (United States)

1991-12-31

Three unusual incidents and case histories involving uranium hexafluoride in the enrichment facilities of the USA in the late 1940`s and early 1950`s are presented. The history of the measurements of the thermal expansion of liquids containing fluorine atoms within the molecule is reviewed with special emphasis upon uranium hexafluoride. A comparison is made between fluorinated esters, fluorocarbons, and uranium hexafluoride. The quantitative relationship between the thermal expansion coefficient, a, of liquids and the critical temperature, T{sub c} is presented. Uranium hexafluoride has an a that is very high in a temperature range that is used by laboratory and production workers - much higher than any other liquid measured. This physical property of UF{sub 6} has resulted in accidents involving filling the UF{sub 6} containers too full and then heating with a resulting rupture of the container. Such an incident at a uranium gaseous diffusion plant is presented. Production workers seldom {open_quotes}see{close_quotes} uranium hexafluoride. The movement of UF{sub 6} from one container to another is usually trailed by weight, not sight. Even laboratory scientists seldom {open_quotes}see{close_quotes} solid or liquid UF{sub 6} and this can be a problem at times. This inability to {open_quotes}see{close_quotes} the UF{sub 6}-HF mixtures in the 61.2{degrees}C to 101{degrees}C temperature range caused a delay in the understanding of the phase diagram of UF{sub 6}-HF which has a liquid - liquid immiscible region that made the eutectic composition somewhat elusive. Transparent fluorothene tubes solved the problem both for the UF{sub 6}-HF phase diagram as well as the UF{sub 6}-HF-CIF{sub 3} phase diagram with a miscibility gap starting at 53{degrees}C. The historical background leading to the first use of CIF{sub 3} to produce UF{sub 6} in both the laboratory and plant at K-25 is presented.

Chemical characterization of SRP waste tank sludges and supernates

International Nuclear Information System (INIS)

Gray, L.W.; Donnan, M.Y.; Okamoto, B.Y.

1979-08-01

Most high-level liquid wastes at the Savannah River Plant (SRP) are byproducts from plutonium and enriched uranium recovery processes. The high-level liquid wastes generated by these separations processes are stored in large, underground, carbon-steel tanks. The liquid wastes consist of: supernate (an aqueous solution containing sodium, nitrate, nitrite, hydroxyl, and aluminate ions), sludge (a gelatinous material containing insoluble components of the waste, such as ferric and aluminum hydroxides, and mercuric and manganese oxides), and salt cake (crystals, such as sodium nitrate, formed by evaporation of water from supernate). Analyses of SRP wastes by laser-Raman spectrometry, atomic absorption spectrometry, spark-source mass spectrometry, neutron activation analysis, colorimetry, ion chromatography, and various other wet-chemical and radiochemical methods are discussed. These analyses are useful in studies of waste tank corrosion and of forms for long-term waste storage

Process and device for liquid organic waste processing by sulfuric mineralization

International Nuclear Information System (INIS)

Aspart, A.; Gillet, B.; Lours, S.; Guillaume, B.

1990-01-01

In a chemical reactor containing sulfuric acid are introduced the liquid waste and nitric acid at a controlled flow rate for carbonization of the waste and oxidation of carbon on sulfur dioxide, formed during carbonization, regenerating simultaneously sulfuric acid. Optical density of the liquid is monitored to stop liquid waste feeding above a set-point. The liquid waste can be an organic solvent such as TBP [fr

Method of concentrating radioactive liquid waste

International Nuclear Information System (INIS)

Yasumura, Keijiro

1990-01-01

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

Containment of uranium in the proposed Egyptian geologic repository for radioactive waste using hydroxyapatite

International Nuclear Information System (INIS)

Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Headley, Thomas Jeffrey; Sanchez, Charles Anthony; Zhao, Hongting; Salas, Fred Manuel; Hasan, Mahmoud A.; Holt, Kathleen Caroline

2003-01-01

Currently, the Egyptian Atomic Energy Authority is designing a shallow-land disposal facility for low-level radioactive waste. To insure containment and prevent migration of radionuclides from the site, the use of a reactive backfill material is being considered. One material under consideration is hydroxyapatite, Ca 10 (PO 4 ) 6 (OH) 2 , which has a high affinity for the sorption of many radionuclides. Hydroxyapatite has many properties that make it an ideal material for use as a backfill including low water solubility (K sp > 10 -40 ), high stability under reducing and oxidizing conditions over a wide temperature range, availability, and low cost. However, there is often considerable variation in the properties of apatites depending on source and method of preparation. In this work, we characterized and compared a synthetic hydroxyapatite with hydroxyapatites prepared from cattle bone calcined at 500 C, 700 C, 900 C and 1100 C. The analysis indicated the synthetic hydroxyapatite was similar in morphology to 500 C prepared cattle hydroxyapatite. With increasing calcination temperature the crystallinity and crystal size of the hydroxyapatites increased and the BET surface area and carbonate concentration decreased. Batch sorption experiments were performed to determine the effectiveness of each material to sorb uranium. Sorption of U was strong regardless of apatite type indicating all apatite materials evaluated. Sixty day desorption experiments indicated desorption of uranium for each hydroxyapatite was negligible.

Containment of uranium in the proposed Egyptian geologic repository for radioactive waste using hydroxyapatite

International Nuclear Information System (INIS)

Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Headley, Thomas Jeffrey; Sanchez, Charles Anthony; Zhao, Hongting; Salas, Fred Manuel; Hasan, Mahmoud A.; Holt, Kathleen Caroline

2004-01-01

Currently, the Egyptian Atomic Energy Authority is designing a shallow-land disposal facility for low-level radioactive waste. To insure containment and prevent migration of radionuclides from the site, the use of a reactive backfill material is being considered. One material under consideration is hydroxyapatite, Ca 10 (PO 4 ) 6 (OH) 2 , which has a high affinity for the sorption of many radionuclides. Hydroxyapatite has many properties that make it an ideal material for use as a backfill including low water solubility (K sp >10 -40 ), high stability under reducing and oxidizing conditions over a wide temperature range, availability, and low cost. However, there is often considerable variation in the properties of apatites depending on source and method of preparation. In this work, we characterized and compared a synthetic hydroxyapatite with hydroxyapatites prepared from cattle bone calcined at 500 C, 700 C, 900 C and 1100 C. The analysis indicated the synthetic hydroxyapatite was similar in morphology to 500 C prepared cattle hydroxyapatite. With increasing calcination temperature the crystallinity and crystal size of the hydroxyapatites increased and the BET surface area and carbonate concentration decreased. Batch sorption experiments were performed to determine the effectiveness of each material to sorb uranium. Sorption of U was strong regardless of apatite type indicating all apatite materials evaluated. Sixty day desorption experiments indicated desorption of uranium for each hydroxyapatite was negligible

Waste container and method for containing waste

International Nuclear Information System (INIS)

Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

1990-01-01

In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

Container for processing and disposing radioactive wastes and industrial wastes

International Nuclear Information System (INIS)

Araki, Kunio; Kasahara, Yuko; Kasai, Noboru; Sudo, Giichi; Ishizaki, Kanjiro.

1978-01-01

Purpose: To improve the performance of containers for radioactive wastes for ocean disposal and on-land disposal such as impact strength, chemical resistance, fire resistance, corrosion resistance, water impermeability and the like. Constitution: Steel fiber-reinforced concrete previously molded in a shape of a container is impregnated with polymerizable impregnating agent selected from the group consisting of a polymerizable monomer, liquid mixture of a polymerizable monomer and an oligomer, a polymer solution, a copolymer solution and the liquid mixture thereof. Then, the polymerizable impregnating agent is polymerized to solidify in the concrete by way of heat-polymerization or radiation-induced polymerization to form a waste container. The container thus obtained can be improved with the impact resistance and wear resistance and further improved with salt water resistance, acid resistance, corrosion resistance and solidity by the impregnation of the polymer, as well as can effectively be prevented from leaching out of radioactive substances. (Furukawa, Y.)

Disposal and reclamation of southwestern coal and uranium wastes

International Nuclear Information System (INIS)

Wewerka, E.M.

1979-01-01

The types of solid wastes and effluents produced by the southwestern coal and uranium mining and milling industries are considered, and the current methods for the disposal and reclamation of these materials discussed. The major means of disposing of the solid wastes from both industries is by land fill or in some instances ponding. Sludges or aqueous wastes are normally discharged into settling and evaporative ponds. Basic reclamation measures for nearly all coal and uranium waste disposal sites include solids stabilization, compacting, grading, soil preparation, and revegetation. Impermeable liners and caps are beginning to be applied to disposal sites for some of the more harmful coal and uranium waste materials

The Assessment of Radioactive Liquid Waste Treatment Generated From The Fuel Reprocessing Plant Using Chemical Coagulation Method

International Nuclear Information System (INIS)

Kuncoro Arief, H; M Birmano, Dj

1998-01-01

Reprocessing of nuclear spent fuel produced 8 lot of radioactive liquid waste still bearing uranium and transuranium. The assessment of the radioactive liquid waste treatment with FeCI 3 as coagulant has been done. Decontamination factor and separation efficiency can be calculated from known activities of initial and post-treatment wastes. It can be concluded that some factors i.e. pH of treatment process, quantity of coagulant, mixing rate, and mixing time have influenced the treatment product

Uranium conversion wastes

International Nuclear Information System (INIS)

Vicente, R.; Dellamano, J.C.

1989-12-01

A set of mathematical equations was developed and used to estimate the radiological significance of each radionuclide potentially present in the uranium refining industry effluents. The equations described the evolution in time of the radionuclides activities in the uranium fuel cycle, from mining and milling, through the yellowcake, till the conversion effluents. Some radionuclides that are not usually monitored in conversion effluents (e.g. Pa-231 and Ac-227) were found to be potentially relevant from the radiological point of view in conversion facilities, and are certainly relevant in mining and milling industry, at least in a few waste streams. (author) [pt

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

Saslow, Sarah A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Asmussen, Robert M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sahajpal, Rahul [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-07-01

This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. ⁹⁹Tc and ¹²⁹I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.

Demonstration of pyropartitioning process by using genuine high-level liquid waste. Reductive-extraction of actinide elements from chlorination product

International Nuclear Information System (INIS)

Uozumi, Koichi; Iizuka, Masatoshi; Kurata, Masaki; Ougier, Michel; Malmbeck, Rikard; Winckel, Stefaan van

2009-01-01

The pyropartitioning process separates the minor actinide elements (MAs) together with uranium and plutonium from the high-level liquid waste generated at the Purex reprocessing of spent LWR fuel and introduces them to metallic fuel cycle. For the demonstration of this technology, a series experiment using 520g of genuine high-level liquid waste was started and the conversion of actinide elements to their chlorides was already demonstrated by denitration and chlorination. In the present study, a reductive extraction experiment in molten salt/liquid cadmium system to recover actinide elements from the chlorination product of the genuine high-level liquid waste was performed. The results of the experiment are as following; 1) By the addition of the cadmium-lithium alloy reductant, almost all of plutonium and MAs in the initial high-level liquid waste were recovered in the cadmium phase. It means no mass loss during denitration, chlorination, and reductive-extraction. 2) The separation factor values of plutonium, MAs, and rare-earth fission product elements versus uranium agreed with the literature values. Therefore, actinide elements will be separated from fission product elements in the actual system. Hence, the pyropartitioning process was successfully demonstrated. (author)

Treatment of liquid waste containing alpha nuclides by adsorption

International Nuclear Information System (INIS)

Zeng Jishu; Su Xiguang; Xia Dejing; Fan Sianhua

1997-01-01

In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10 3 Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs

Bioaccumulation of uranium from waste water using different strains of Saccharomyces cerevisiae

International Nuclear Information System (INIS)

Tykva, R.; Novak, J.; Podracka, E.; Popa, K.

2009-01-01

Five different strains of Saccharomyces cerevisiae were tested for their abilities to accumulate uranium from waste water containing competitive ions. Samples of water passing out from a previous uranium mill were used. The strains tested possess different abilities to accumulate uranium. The kinetics of bioaccumulation, the leaching degree, the influence of cell density and their origin were investigated. Under the applied experimental conditions, more than a half of the total activity (uranium and the decay products) could be accumulated after 60 min contact time of 1 mL (S. cerevisiae) suspension and 5 mL of water. The other cations present in solution effectively competed for the uranium accumulation. 226 Ra and its decay products were completely retained using all tested strains. (authors)

Problems in the design and specification of containers for vitrified high-level liquid waste

International Nuclear Information System (INIS)

Corbet, A.D.W.; Hall, G.G.; Spiller, G.T.

1976-01-01

In the United Kingdom the growing problem of ensuring the safe storage of high-level liquid waste over long time scales has led to a policy for implementing solidification. A brief description is given of the HARVEST vitrification process, which is essentially a scaled-up version of the FINGAL process with increased throughput. The functional requirements of the container are considered. It must be made of a material which can be fabricated to a high standard. Diameters up to 600 mm for right circular cylindrical containers and 1200 mm for annular containers are contemplated. Computer aids for axisymmetric and three-dimensional heat transfer and stress analysis are identified. One example is given of the thermal profile for the cylindrical container in the furnace and another example for the annular container following an accident condition. Measured values are given for high temperature oxidation, emissivity and the short-term creep strength of various alloys. Corrosion in fresh water and sea water over long time periods and leaching of partially exposed solid waste are discussed and a conceptual package for sea bed disposal is described. The relative merits of the different methods of manufacture are pointed out and the paper concludes that HK-40 or better INCOLOY alloy 800L are suitable materials of construction. (author)

National Public Information Symposium on Peaceful Uses of Nuclear Energy, NUC Info'2000. Radioactive Waste Management and Site Restoration in Uranium Industry. Proceedings. Volume 1

International Nuclear Information System (INIS)

Dobos, Ion; Comsa, Olivia

2000-01-01

These proceedings published in two volumes contain materials presented at the National Public Information Symposium on Peaceful Uses of Nuclear Energy, NUC Info'2000, Radioactive Waste Management and Site Restoration in Uranium Industry, held on 5. September to 8. September 2000 at Baita Bihor, Romania. As the name of Symposium indicates, this manifestation is addressed not only to specialists but rather to the public at large. The proceedings are structured in 4 sections: 1. Management of radioactive waste arising from uranium mining, milling and decommissioning; 2. Uranium mine close-down; 3. Environmental restoration of uranium mining and milling sites; 4. Management of radioactive waste arising from nuclear applications. The first volume also contains an inaugural session dedicated to nuclear power, nuclear fuel cycle and development of uranium industry in Romania. The contributions in the first volume deal with the management of radioactive waste arising from uranium mining, milling and decommissioning and uranium mine close-out

The Study of Isolated Bacteria Application for Bioremediation Agent of Uranium Radionuclide in the Environment

International Nuclear Information System (INIS)

Yazid, Mochd

2007-01-01

Application of the isolated bacteria on the Low Level Uranium Waste as uranium bioremediation agent in the environment has been studied. The objective of this research is to study the possibility of isolated bacteria to be used on uranium remediation process. The isolation of uranium resistance bacteria was carried out on the selective medium SBS containing 10 mg/l uranium, incubated at 37°C until the growth was visible. Selection of binding uranium bacteria was carried out based on their ability to grow on liquid medium containing various concentration of uranium that shacked on 120 rpm speed. The isolated bacteria with the highest specific growth rate constant (μ) were selected for biochemical characterization and identification by matching profile method. The result of this research showed that three selected isolate bacteria were able to grow well on liquid SBS medium until 100 mg/l uranium concentration. The identification results showed that two of them were suspected belong to the genus Pseudomonas and one isolates belong to the genus of Bacillus. The uranium reduction studied was performed by growing up the isolated bacteria on the SBS liquid medium that containing 40 mg/l uranium. Bacterial growth were measured by weighted of bacterial biomass and uranium concentration were measured by spectrophotometer. The research result showed that the selected isolates bacteria may applicable for bioremediation agent because of their ability to grow well on liquid SBS medium and their ability on uranium concentration reduction. The efficiency of reduction by Pseudomonas in the isolated bacteria one were 78.51 % and in the isolated bacteria three were 91.47 % , and Bacillus in the isolate bacteria six were 52.73%. (author)

Liquid waste sampling device

International Nuclear Information System (INIS)

Kosuge, Tadashi

1998-01-01

A liquid pumping pressure regulator is disposed on the midway of a pressure control tube which connects the upper portion of a sampling pot and the upper portion of a liquid waste storage vessel. With such a constitution, when the pressure in the sampling pot is made negative, and liquid wastes are sucked to the liquid pumping tube passing through the sampling pot, the difference between the pressure on the entrance of the liquid pumping pressure regulator of the pressure regulating tube and the pressure at the bottom of the liquid waste storage vessel is made constant. An opening degree controlling meter is disposed to control the degree of opening of a pressure regulating valve for sending actuation pressurized air to the liquid pumping pressure regulator. Accordingly, even if the liquid level of liquid wastes in the liquid waste storage vessel is changed, the height for the suction of the liquid wastes in the liquid pumping tube can be kept constant. With such procedures, sampling can be conducted correctly, and the discharge of the liquid wastes to the outside can be prevented. (T.M.)

Physicochemical aspects of extraction of uranium concentrate from the wastes and thermodynamic characteristics of thorium-uranium compounds

International Nuclear Information System (INIS)

Khamidov, F.A.

2017-01-01

The purpose of present work is elaboration of physicochemical aspects of extraction of uranium concentrate from the wastes and study of thermodynamic characteristics of thorium-uranium compounds. Therefore, the radiological monitoring of tailing dumps of Tajikistan has been conducted; the obtaining of uranium concentrate from the tailing dumps of uranium production has been studied; the obtaining of uranium concentrate from the tailing dumps of uranium production with application of local sorbents has been studied as well; thermal stability and thermodynamic characteristics of uranium-thorium compounds has been investigated; the flowsheets of extraction of uranium concentrate from the wastes have been elaborated.

Reduction of radioactive waste from remediation of uranium-contaminated soil

Energy Technology Data Exchange (ETDEWEB)

Kim, Il Gook; Kim, Seung Soo; Kim, Gye Nam; Han, Gyu Seong; Choi, Jong Won [Decontamination and Decommissioning Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-06-15

Great amounts of solid radioactive waste (second waste) and waste solution are generated from the remediation of uranium-contaminated soil. To reduce these, we investigated washing with a less acidic solution and recycling the waste solution after removal of the dominant elements and uranium. Increasing the pH of the washing solution from 0.5 to 1.5 would be beneficial in terms of economics. A high content of calcium in the waste solution was precipitated by adding sulfuric acid. The second waste can be significantly reduced by using sorption and desorption techniques on ampholyte resin S-950 prior to the precipitation of uranium at pH 3.0.

Reduction of radioactive waste from remediation of uranium-contaminated soil

International Nuclear Information System (INIS)

Kim, Il Gook; Kim, Seung Soo; Kim, Gye Nam; Han, Gyu Seong; Choi, Jong Won

2016-01-01

Great amounts of solid radioactive waste (second waste) and waste solution are generated from the remediation of uranium-contaminated soil. To reduce these, we investigated washing with a less acidic solution and recycling the waste solution after removal of the dominant elements and uranium. Increasing the pH of the washing solution from 0.5 to 1.5 would be beneficial in terms of economics. A high content of calcium in the waste solution was precipitated by adding sulfuric acid. The second waste can be significantly reduced by using sorption and desorption techniques on ampholyte resin S-950 prior to the precipitation of uranium at pH 3.0

Chlorination of antimony and its volatilization treatment of waste antimony-uranium composite oxide catalyst

International Nuclear Information System (INIS)

Sawada, K.; Enokida, Y.

2011-01-01

For the waste antimony-uranium composite oxide catalyst, the chlorination of antimony and its volatilization treatment were proposed, and evaluated using hydrogen chloride gas at 873-1173 K. During the treatment, the weight loss of the composite oxide sample, which resulted from the volatilization of antimony, was confirmed. An X-ray diffraction analysis showed that uranium oxide, U 3 O 8 , was formed during the reaction. After the treatment at 1173 K for 1 h, almost all the uranium contained in the waste catalyst was dissolved by a 3 M nitric acid solution at 353 K within 10 min, although that of the non-treated catalyst was less than 0.1%. It was found that the chlorination and volatilization treatment was effective to separate antimony from the composite oxide catalyst and change uranium into its removable form. (orig.)

The Treatment of Mixed Waste with GeoMelt In-Container Vitrification

International Nuclear Information System (INIS)

Finucane, K.G.; Campbell, B.E.

2006-01-01

AMEC's GeoMelt R In-Container Vitrification (ICV) TM has been used to treat diverse types of mixed low-level radioactive waste. ICV is effective in the treatment of mixed wastes containing polychlorinated biphenyls (PCBs) and other semi-volatile organic compounds, volatile organic compounds (VOCs) and heavy metals. The GeoMelt vitrification process destroys organic compounds and immobilizes metals and radionuclides in an extremely durable glass waste form. The process is flexible allowing for treatment of aqueous, oily, and solid mixed waste, including contaminated soil. In 2004, ICV was used to treat mixed radioactive waste sludge containing PCBs generated from a commercial cleanup project regulated by the Toxic Substances Control Act (TSCA), and to treat contaminated soil from Rocky Flats Environmental Technology Site. The Rocky Flats soil contained cadmium, PCBs, and depleted uranium. In 2005, AMEC completed a treatability demonstration of the ICV technology on Mock High Explosive from Sandia National Laboratories. This paper summarizes results from these mixed waste treatment projects. (authors)

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)

High-sensitive detection by direct interrogation of 14 MeV Acc neutrons, (1). Uranium-contained metal matrix in a waste dram

International Nuclear Information System (INIS)

Haruyama, Mitsuo; Takase, Misao; Tobita, Hiroshi; Mori, Takamasa

2004-01-01

Previously, authors reported that the 14 MeV-neutron direct interrogation method has made possible measure for the discrimination of clearance levels of concrete solidification uranium waste. In this paper, applicability of the method to metal waste matrix is discussed based on the results of simulation experiments by the continuation energy Monte Carlo calculation code (MVP). The problem is that self-neutron moderation effect in a waste cannot be expected when a waste matrix is metal. To solve this, a moderator is adopted so as to surround a metal waste drum and to slow down suitably a 14 MeV neutrons. The simulation calculation showed that this effect is satisfactorily large. The detection limit of radioactivity concentration to 4.5% enriched uranium has been found to be 0.0973 Bq/g in the metal waste model of 215.59 kg gross weight, in which 61 pipes are stuffed into its drum. Moreover, the position-dependent sensitivity difference in a metal waste drum can be settled as small as to ±13.5%. In conclusion, it can be said that 14 MeV-neutron direct interrogation method can be applied to the waste of a metal system: the detection sensitivity is high enough and the position-dependent sensitivity difference is small admittedly. Hence the method can be applied also to discrimination measurement of the clearance level of metal uranium waste. (author)

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

International Nuclear Information System (INIS)

Rubin, L.S.

1986-01-01

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

Treatment of liquid waste containing alpha nuclides by adsorption

Energy Technology Data Exchange (ETDEWEB)

Jishu, Zeng; Xiguang, Su; Dejing, Xia; Sianhua, Fan [China Inst. of Atomic Energy, Beijing (China). Radiochemistry Dept.

1997-02-01

In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10{sup 3} Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs.

Method for solidifying liquid radioactive wastes

International Nuclear Information System (INIS)

Berreth, J.R.

1976-01-01

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

Radioactive liquid waste filtering device

International Nuclear Information System (INIS)

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

1988-01-01

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

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

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

Supercritical fluid extraction of uranium

International Nuclear Information System (INIS)

Kumar, Pradeep

2017-01-01

Uranium being strategic material, its separation and purification is of utmost importance in nuclear industry, for which solvent extraction is being employed. During solvent extraction significant quantity of radioactive liquid waste gets generated which is of environmental concern. In recent decades supercritical fluid extraction (SFE) has emerged as promising alternative to solvent extraction owing to its inherent advantage of reduction in liquid waste generation and simplification of process. In this paper a brief overview of research work carried out so far on SFE of uranium by BARC has been given

Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

International Nuclear Information System (INIS)

Ferreira, Rafael Vicente de Padua

2008-01-01

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

The Treatment of Low Level Radioactive Liquid Waste Containing Detergent by Biological Activated Sludge Process

International Nuclear Information System (INIS)

Zainus Salimin

2002-01-01

The treatment of low level radioactive liquid waste containing persil detergent from laundry operation of contaminated clothes by activated sludge process has been done, for alternative process replacing the existing treatment by evaporation. The detergent concentration in water solution from laundry operation is 14.96 g/l. After rinsing operation of clothes and mixing of laundry water solution with another liquid waste, the waste water solution contains about ≤ 1.496 g/l of detergent and 10 -3 Ci/m 3 of Cs-137 activity. The simulation waste having equivalent activity of Cs-137 10 -3 Ci/m 3 , detergent content (X) 1.496, 0.748, 0.374, 0.187, 0.1496 and 0.094 g/l on BOD value respectively 186, 115, 71, 48, 19, and 16 ppm was processed by activated sludge in reactor of 18.6 l capacity on ambient temperature. It is used Super Growth Bacteria (SGB) 102 and SGB 104, nitrogen and phosphor nutrition, and aeration. The result show that bacteria of SGB 102 and SGB 104 were able to degrade the persil detergent for attaining standard quality of water release category B in which BOD values 6 ppm. It was need 30 hours for X ≤ 0.187 g/l, 50 hours for 0.187 < X ≤ 0.374 g/l, 75 hours for 0.374 < X ≤ 0.748, and 100 hours for 0.748 < X ≤ 1.496 g/l. On the initial period the bacteria of SGB 104 interact most quickly to degrade the detergent comparing SGB 102. Biochemical oxidation process decontaminate the solution on the decontamination factor of 350, Cs-137 be concentrate in sludge by complexing with the bacteria wall until the activity of solution be become very low. (author)

Radiolytic decomposition of dioxins in liquid wastes

International Nuclear Information System (INIS)

Zhao Changli; Taguchi, M.; Hirota, K.; Takigami, M.; Kojima, T.

2006-01-01

The dioxins including polychlorinated dibenzo-p-dioxins (PCDDs) and polychlorinated dibenzofurans (PCDFs) are some of the most toxic persistent organic pollutants. These chemicals have widely contaminated the air, water, and soil. They would accumulate in the living body through the food chains, leading to a serious public health hazard. In the present study, radiolytic decomposition of dioxins has been investigated in liquid wastes, including organic waste and waste-water. Dioxin-containing organic wastes are commonly generated in nonane or toluene. However, it was found that high radiation doses are required to completely decompose dioxins in the two solvents. The decomposition was more efficient in ethanol than in nonane or toluene. The addition of ethanol to toluene or nonane could achieve >90% decomposition of dioxins at the dose of 100 kGy. Thus, dioxin-containing organic wastes can be treated as regular organic wastes after addition of ethanol and subsequent γ-ray irradiation. On the other hand, radiolytic decomposition of dioxins easily occurred in pure-water than in waste-water, because the reaction species is largely scavenged by the dominant organic materials in waste-water. Dechlorination was not a major reaction pathway for the radiolysis of dioxin in water. In addition, radiolytic mechanism and dechlorinated pathways in liquid wastes were also discussed. (authors)

Management of wastes from uranium mines and mills

International Nuclear Information System (INIS)

Thomas, K.T.

1981-01-01

Uranium mining and milling operations have not given rise to much concern about their hazards, and with advancing technologies for mill processing and waste management, the situation will continue to improve. However, the disposal of large quantities of waste produced in mining and milling does have an environmental impact, owing to the long half-lives and the ready availability of the toxic radionuclides Ra-226 and Rn-222. This article deals with the management of wastes from uranium mines and mills

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

International Nuclear Information System (INIS)

Morais, Carlos Antonio de

1996-01-01

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

Chemical thermodynamics of uranium

International Nuclear Information System (INIS)

Grenthe, I.; Fuger, J.; Lemire, R.J.; Muller, A.B.; Nguyen-Trung Cregu, C.; Wanner, H.

1992-01-01

A comprehensive overview on the chemical thermodynamics of those elements that are of particular importance in the safety assessment of radioactive waste disposal systems is provided. This is the first volume in a series of critical reviews to be published on this subject. The book provides an extensive compilation of chemical thermodynamic data for uranium. A description of procedures for activity corrections and uncertainty estimates is given. A critical discussion of data needed for nuclear waste management assessments, including areas where significant gaps of knowledge exist is presented. A detailed inventory of chemical thermodynamic data for inorganic compounds and complexes of uranium is listed. Data and their uncertainty limits are recommended for 74 aqueous complexes and 199 solid and 31 gaseous compounds containing uranium, and on 52 aqueous and 17 solid auxiliary species containing no uranium. The data are internally consistent and compatible with the CODATA Key Values. The book contains a detailed discussion of procedures used for activity factor corrections in aqueous solution, as well as including methods for making uncertainty estimates. The recommended data have been prepared for use in environmental geochemistry. Containing contributions written by experts the chapters cover various subject areas such a s: oxide and hydroxide compounds and complexes, the uranium nitrides, the solid uranium nitrates and the arsenic-containing uranium compounds, uranates, procedures for consistent estimation of entropies, gaseous and solid uranium halides, gaseous uranium oxides, solid phosphorous-containing uranium compounds, alkali metal uranates, uncertainties, standards and conventions, aqueous complexes, uranium minerals dealing with solubility products and ionic strength corrections. The book is intended for nuclear research establishments and consulting firms dealing with uranium mining and nuclear waste disposal, as well as academic and research institutes

Potential health hazard of nuclear fuel waste and uranium ore

International Nuclear Information System (INIS)

Mehta, K.; Sherman, G.R.; King, S.G.

1991-06-01

The variation of the radioactivity of nuclear fuel waste (used fuel and fuel reprocessing waste) with time, and the potential health hazard (or inherent radiotoxicity) resulting from its ingestion are estimated for CANDU (Canada Deuterium Uranium) natural-uranium reactors. Four groups of radionuclides in the nuclear fuel waste are considered: actinides, fission products, activation products of zircaloy, and activation products of fuel impurities. Contributions from each of these groups to the radioactivity and to the potential health hazard are compared and discussed. The potential health hazard resulting from used fuel is then compared with that of uranium ore, mine tailings and refined uranium (fresh fuel) on the basis of equivalent amounts of uranium. The computer code HAZARD, specifically developed for these computations, is described

Control and prevention of seepage from uranium mill waste disposal facilities

International Nuclear Information System (INIS)

Williams, R.E.

1978-01-01

This paper constitutes an analysis of the technologies which are available for the prevention of movement of waste waters out of uranium mill waste disposal facilities via sub-surface routes. Hydrogeologic criteria for potential uranium mill waste disposal sites and mathematical modeling of contaminant migration in ground water are presented. Methods for prevention of seepage from uranium mill waste disposal facilities are investigated: liners, clay seals, synthetic polymeric membranes (PVC, polyethylene, chlorinated polyethylene, hypalon, butyl rubber, neoprene, elasticized polyolefin)

Differential lead retention in zircons: implications for nuclear waste containment.

Science.gov (United States)

Gentry, R V; Sworski, T J; McKown, H S; Smith, D H; Eby, R E; Christie, W H

1982-04-16

An innovative ultrasensitive technique was used for lead isotopic analysis of individual zircons extracted from granite core samples at depths of 960, 2170, 2900, 3930, and 4310 meters. The results show that lead, a relatively mobile element compared to the nuclear waste-related actinides uranium and thorium, has been highly retained at elevated temperatures (105 degrees to 313 degrees C) under conditions relevant to the burial of synthetic rock waste containers in deep granite holes.

Alternative repository criticality-control strategies for fissile uranium wastes

International Nuclear Information System (INIS)

Forsberg, C.W.

1998-01-01

Methods to prevent long term, disposal site nuclear criticality from fissile uranium isotopes in wastes were investigated. Long term refers to the time period after waste package (WP) failure and the subsequent loss of geometry and chemistry control within the WP. The preferred method of control was found to be the addition of sufficient depleted uranium to each WP so that the uranium enrichment is reduced to 235 U and 233 U in 238 U

Management of wastes from the refining and conversion of uranium ore concentrate to uranium hexafluoride

International Nuclear Information System (INIS)

1981-01-01

This report is the outcome of an IAEA Advisory Group Meeting on ''Waste Management Aspects in Relation to the Refining of Uranium Ore Concentrates and their Conversion to Uranium Hexafluoride'', which was held in Vienna from 17 to 21 December 1979. The report summarizes the main topics discussed at the meeting and gives an overview of uranium refining processes, being used in nuclear industry. The meeting was organized by the International Atomic Energy Agency, Radioactive Waste Management Section

The radioactive waste management at IAEA laboratories

International Nuclear Information System (INIS)

Deron, S.; Ouvrard, R.; Hartmann, R.; Klose, H.

1992-10-01

The report gives a brief description of the nature of the radioactive wastes generated at the IAEA Laboratories in Seibersdorf, their origin and composition, their management and monitoring. The management of the radioactive waste produced at IAEA Laboratories in Seibersdorf is governed by the Technical Agreements of 1985 between the IAEA and the Austrian Health Ministry. In the period of 1982 to 1991 waste containers of low activity and radiotoxicity generated at laboratories other than the Safeguards Analytical Laboratory (SAL) were transferred to the FZS waste treatment and storage plant: The total activity contained in these drums amounted to < 65 MBq alpha activity; < 1030 MBq beta activity; < 2900 MBq gamma activity. The radioactive waste generated at SAL and transferred to the FZs during the same period included. Uranium contaminated solid burnable waste in 200 1 drums, uranium contaminated solid unburnable waste in 200 1 drums, uranium contaminated liquid unburnable waste in 30 1 bottles, plutonium contaminated solid unburnable waste in 200 1 drums. In the same period SAL received a total of 146 Kg uranium and 812 g plutonium and exported out of Austria, unused residues of samples. The balance, i.e.: uranium 39 kg, plutonium 133 g constitutes the increase of the inventory of reference materials, and unused residues awaiting export, accumulated at SAL and SIL fissile store as a result of SAL operation during this 10 year period. The IAEA reexports all unused residues of samples of radioactive and fissile materials analyzed at his laboratories, so that the amount of radioactive materials ending in the wastes treated and stored at FZS is kept to a minimum. 5 refs, 7 figs, 3 tabs

Behaviour of uranium dioxide in liquid nitrogen tetraoxide

International Nuclear Information System (INIS)

Kobets, L.V.; Klavsut', G.N.; Dolgov, V.M.

1983-01-01

Interaction kinetics of uranium dioxide with liquid nitrogen tetroxide at 25-150 deg C has been studied. It is shown that in the temperature range studied NO[UO 2 (NO 3 ) 3 ] is the final product of the reaction. With the increase of specific surface of uranium dioxide and with the temperature increase the degree of oxide transformation increases. Uranium dioxide-liquid N 2 O 4 interaction proceeds in the diffusion region. Seeming activation energies and rate constants of the mentioned processes are calculated. Effect of nitrogen trioxide additions on transformation kinetics is considered

Study of the extraction and the purification of americium and trivalent actinides contained in effluents with supported liquid membranes

International Nuclear Information System (INIS)

Guillou, P.

1990-12-01

The supported liquid membrane technique is studied and developed for americium recovery from uranium or plutonium matrices and decontamination of liquid radioactive wastes. First tests on uranium-nickel solutions with a flat membrane showed the easiness of the operation and the efficiency of the process. Acid-resistant (10 N), interchangeable elements with hollow fibers, are developed and also a computerized automatic device. The different tests on americium solutions demonstrate the feasibility and the reliability of the system. Influence of various parameters on transfer kinetics is investigated

Radioactive Waste Issues related to Production of Fission-based Mo-99 by using Low Enriched Uranium (LEU)

Energy Technology Data Exchange (ETDEWEB)

Hassan, Muhmood ul; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

In order to produce fission-based Mo-99 from research reactors, two types of targets are being used and they are highly enriched uranium (HEU) targets with {sup 235}U enrichment more than 90wt% of {sup 235}U and low enriched uranium (LEU) targets with {sup 235}U enrichment less than 20wt% of {sup 235}U. It is worth noting that medium enriched uranium i.e. 36wt% of {sup 235}U as being used in South Africa is also regarded as non-LEU from a nuclear security point of view. In order to cope with the proliferation issues, international nuclear security policy is promoting the use of LEU targets in order to minimize the civilian use of HEU. It is noteworthy that Mo-99 yield of the LEU target is less than 20% of the HEU target, which requires approximately five times more LEU targets to be irradiated and consequently results in increased volume of waste. The waste generated from fission Mo-99 production can be mainly due to: target fabrication, assembling of target, irradiation in reactor and processing of irradiated targets. During the fission of U-235 in a reactor, a large number of radionuclides with different chemical and physical properties are formed. The waste produced from these practices may be a combination of low level waste (LLW) and intermediate level waste (ILW) comprised of all three types, i.e., solid, liquid and gas. Handling and treatment of the generated waste are dependent on its form and activity. In case of the large production facility, waste storage facility should be constructed in order to limit the radiation exposures of the workers and the environment. In this study, we discuss and compare mainly the radioactive waste generated by alkaline digestion of both HEU and LEU targets to assist in planning and deciding the choice of the technology with better arrangements for proper handling and disposal of generated waste. With the use of the LEU targets in Mo-99 production facility, significant increase in liquid and solid waste has been expected.

Interactions of low-level, liquid radioactive wastes with soils. 1. Behavior of radionuclides in soil-waste systems

International Nuclear Information System (INIS)

Fowler, E.B.; Essington, E.H.; Polzer, W.L.

1981-01-01

The characteristics of radioactive wastes and soils vary over a wide range. Liquid radioactive waste entering the environment will eventually contact the soil or geological matrix; interactions will be determined by the chemical and physical nature of the liquid, as well as the soil matrix. We report here the results from an investigation of certain of those characteristics as they relate to retention of radionuclides by soils. Three fractions were demonstrated in the waste as filterable, soluble-sorbable, and soluble-nonsorbable; the physical nature of each fraction was demonstrated using autoradiographic techniques. Isotopes of plutonium and uranium and americium-241 in the soluble fraction of the waste were shown to have a negative charge as determined by ion exchange techniques. In the soil-waste systems, the net charge for those radionuclides was shown to change from predominantly negative to predominantly positive. Nevertheless, cesium-137 was shown to be predominantly positited by TVA and approved by NRC (formerly AEC) since June 1973. This report is based upon the revisions, approved through the end of this reporting period

Challenges in waste management and environmental restoration in the uranium mining industry

International Nuclear Information System (INIS)

Jarrell, J.

2011-01-01

Two components dominate the waste management efforts at conventional Canadian uranium mining and milling operations. These are the waste rock generated in the mining of ore as well as the mill tailings -- which are the residue solids remaining after uranium extraction. Much has changed in the management of these wastes over the years. Visually, current sites are generally more compact than those developed earlier, due to higher grade ores and less land disturbance. However, the more significant strides being made to better manage uranium mining wastes deal more with improved chemical and physical controls rather than those changes which are visible. Segregation of waste rock to separate out potentially problematic material within the more weakly mineralized halo surrounding the ore is now a core strategy. This segregation is based on both the waste rock's chemical and radiological characteristics. Better controls have also been introduced on tailings physical properties to minimize their permeability, along with better chemical controls to minimize tailings contaminant solubility. Efforts to engineer tailings properties are coupled with contrasting hydraulic conductivity between the consolidated tailings mass and surrounding geologic materials. This creates the necessary long-term containment controls built into modern tailings management facilities. Current challenges include selecting the correct decommissioning assumptions such as future land use and required environmental acceptance criteria, along with decisions as to when to carry out reclamation work in the life cycle of the mine and mill. Public discussion of restoration plans throughout the life of the facility is essential to build acceptable solutions. Along with challenges come successes. Most recently, improvements have been made in reducing treated water molybdenum and selenium levels. Other successes include the application of reverse osmosis technology on a large scale, recycling of uranium

Challenges in waste management and environmental restoration in the uranium mining industry

Energy Technology Data Exchange (ETDEWEB)

Jarrell, J. [Cameco Corp., Saskatoon, SK (Canada)

2011-07-01

Two components dominate the waste management efforts at conventional Canadian uranium mining and milling operations. These are the waste rock generated in the mining of ore as well as the mill tailings -- which are the residue solids remaining after uranium extraction. Much has changed in the management of these wastes over the years. Visually, current sites are generally more compact than those developed earlier, due to higher grade ores and less land disturbance. However, the more significant strides being made to better manage uranium mining wastes deal more with improved chemical and physical controls rather than those changes which are visible. Segregation of waste rock to separate out potentially problematic material within the more weakly mineralized halo surrounding the ore is now a core strategy. This segregation is based on both the waste rock's chemical and radiological characteristics. Better controls have also been introduced on tailings physical properties to minimize their permeability, along with better chemical controls to minimize tailings contaminant solubility. Efforts to engineer tailings properties are coupled with contrasting hydraulic conductivity between the consolidated tailings mass and surrounding geologic materials. This creates the necessary long-term containment controls built into modern tailings management facilities. Current challenges include selecting the correct decommissioning assumptions such as future land use and required environmental acceptance criteria, along with decisions as to when to carry out reclamation work in the life cycle of the mine and mill. Public discussion of restoration plans throughout the life of the facility is essential to build acceptable solutions. Along with challenges come successes. Most recently, improvements have been made in reducing treated water molybdenum and selenium levels. Other successes include the application of reverse osmosis technology on a large scale, recycling of uranium

Bioprocessing of a stored mixed liquid waste

Energy Technology Data Exchange (ETDEWEB)

Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

1995-12-31

This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

Characterization of uranium in bituminized radioactive waste drums by self-induced X-ray fluorescence

International Nuclear Information System (INIS)

Pin, Patrick; Perot, Bertrand

2013-06-01

This paper reports the experimental qualification of an original uranium characterization method based on fluorescence X rays induced by the spontaneous gamma emission of bituminized radioactive waste drums. The main 661.7 keV gamma ray following the 137 Cs decay produces by Compton scattering in the bituminized matrix an intense photon continuum around 100 keV, i.e. in the uranium X-ray fluorescence region. 'Self-induced' X-rays produced without using an external source allow a quantitative assessment of uranium as 137 Cs and uranium are homogeneously mixed and distributed in the bituminized matrix. The paper presents the experimental qualification of the method with real waste drums, showing a detection limit well below 1 kg of uranium in 20 min acquisitions while the usual gamma rays of 235 U (185 keV) or 238 U (1001 keV of 234m Pa in the radioactive decay chain) are not detected. The relative uncertainty on the uranium mass assessed by self-induced X-ray fluorescence (SXRF) is about 50%, with a 95% confidence level, taking into account the correction of photon attenuation in the waste matrix. This last indeed contains high atomic numbers elements like uranium, but also barium, in quantities which are not known for each drum. Attenuation is estimated thanks to the peak-to-Compton ratio to limit the corresponding uncertainty. The SXRF uranium masses measured in the real drums are in good agreement with long gamma-ray spectroscopy measurements (1001 keV peak) or with radiochemical analyses. (authors)

Deactivation of waste waters in the Czechoslovak Uranium Industry

International Nuclear Information System (INIS)

Priban, V.

1978-01-01

Deactivation techniques are described used for the treatment of waste waters from uranium mines and uranium chemical treatment plants. With treatment plant waters this is done either by precipitation of radium with barium sulfate or using multistage evaporating units. Mine waste waters are deactivated by sorption on ion exchangers; strongly basic anion exchangers, mostly Wofatit SBW, Varion AP or Ostion AU are used for uranium, while the strongly acidic Ostion KS is used for radium. (Z.M.)

Method and apparatus for glass solidification porcessing for radioactive liquid waste

International Nuclear Information System (INIS)

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

1989-01-01

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

Application of biosorbents in treatment of the radioactive liquid waste; Aplicacao de biossorventes no tratamento de rejeitos radioativos liquidos

Energy Technology Data Exchange (ETDEWEB)

Ferreira, Rafael Vicente de Padua

2014-07-01

Radioactive liquid waste containing organic compounds need special attention, because the treatment processes available are expensive and difficult to manage. The biosorption is a potential treatment technique that has been studied in simulated wastes. The biosorption term is used to describe the removal of metals, non-metals and/or radionuclides by a material from a biological source, regardless of its metabolic activity. Among the potential biomasses, agricultural residues have very attractive features, as they allow for the removal of radionuclides present in the waste using a low cost biosorbent. The aim of this study was to evaluate the potential use of different biomass originating from agricultural products (coconut fiber, coffee husk and rice husk) in the treatment of real radioactive liquid organic waste. Experiments with these biomass were made including 1) Preparation, activation and characterization of biomasses; 2) Conducting biosorption assays; and 3) Evaluation of the product of immobilization of biomasses in cement. The biomasses were tested in raw and activated forms. The activation was carried out with diluted HNO{sub 3} and NaOH solutions. Biosorption assays were performed in polyethylene bottles, in which were added 10 mL of radioactive waste or waste dilutions in deionized water with the same pH and 2% of the biomass (w/v). At the end of the experiment, the biomass was separated by filtration and the remaining concentration of radioisotopes in the filtrate was determined by ICP-OES and gamma spectrometry. The studied waste contains natural uranium, americium-241 and cesium-137. The adopted contact times were 30 min, 1, 2 and 4 hours and the concentrations tested ranged between 10% and 100%. The results were evaluated by maximum experimental sorption capacity and isotherm and kinetics ternary models. The highest sorption capacity was observed with raw coffee husk, with approximate values of 2 mg/g of U (total), 40 x 10{sup -6} mg/g of Am-241 and

Devoluming method of acidic radioactive liquid waste and processing system therefor

International Nuclear Information System (INIS)

Shirai, Takamori; Honda, Tadahiro

1998-01-01

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

An automated, self-verifying system for monitoring uranium in effluent streams

International Nuclear Information System (INIS)

Reda, R.J.; Pickett, J.L.

1992-01-01

In nuclear facilities such as nuclear fuel fabrication plants, a constant vigil is required to ensure that the concentrations of uranium in process or waste streams do not exceed required specifications. The specifications may be dictated by the process owner, a regulatory agency such as the US Nuclear Regulatory Agency or Environmental Protection Agency, or by criticality safety engineering criteria. Traditionally, uranium monitoring in effluent streams has been accomplished by taking periodic samples of the liquid stream and determining the concentration by chemical analysis. Despite its accuracy, chemical sampling is not timely enough for practical use in continuously flowing systems because of the possibility that a significant quantity of uranium may be discharged between sampling intervals. To completely satisfy regulatory standards, the liquid waste stream must be monitored for uranium on a 100% basis. To this end, an automated, radioisotopic liquid-waste monitoring system was developed by GE Nuclear Energy as an integral part of the uranium conversion and waste recovery operations. The system utilizes passive gamma-ray spectroscopy and is thus a robust, on-line, and nondestructive assay for uranium. The system provides uranium concentration data for process monitoring and assures regulatory compliance for criticality safety. A summary of the principles of system operation, calibration, and verification is presented in this paper

Uranium oxide catalysts: environmental applications for treatment of chlorinated organic waste from nuclear industry.

Science.gov (United States)

Lazareva, Svetlana; Ismagilov, Zinfer; Kuznetsov, Vadim; Shikina, Nadezhda; Kerzhentsev, Mikhail

2018-02-05

Huge amounts of nuclear waste, including depleted uranium, significantly contribute to the adverse environmental situation throughout the world. An approach to the effective use of uranium oxides in catalysts for the deep oxidation of chlorine-containing hydrocarbons is suggested. Investigation of the catalytic activity of the synthesized supported uranium oxide catalysts doped with Cr, Mn and Co transition metals in the chlorobenzene oxidation showed that these catalysts are comparable with conventional commercial ones. Physicochemical properties of the catalysts were studied by X-ray diffraction, temperature-programmed reduction with hydrogen (H 2 -TPR), and Fourier transform infrared spectroscopy. The higher activity of Mn- and Co-containing uranium oxide catalysts in the H 2 -TPR and oxidation of chlorobenzene in comparison with non-uranium catalysts may be related to the formation of a new disperse phase represented by uranates. The study of chlorobenzene adsorption revealed that the surface oxygen is involved in the catalytic process.

Cost savings associated with landfilling wastes containing very low levels of uranium

International Nuclear Information System (INIS)

Boggs, C.J.; Shaddoan, W.T.

1996-01-01

The Paducah Gaseous Diffusion Plant (PGDP) has operated captive landfills (both residential and construction/demolition debris) in accordance with the Commonwealth of Kentucky regulations since the early 1980s. Typical waste streams allowed in these landfills include nonhazardous industrial and municipal solid waste (such as paper, plastic, cardboard, cafeteria waste, clothing, wood, asbestos, fly ash, metals, and construction debris). In July 1992, the U.S. Environmental Protection Agency issued new requirements for the disposal of sanitary wastes in a open-quotes contained landfill.close quotes These requirements were promulgated in the 401 Kentucky Administrative Record Chapters 47 and 48 that became effective 30 June 1995. The requirements for a new contained landfill include a synthetic liner made of high-density polyethylene in addition to the traditional 1-meter (3-foot) clay liner and a leachate collection system. A new landfill at Paducah would accept waste streams similar to those that have been accepted in the past. The permit for the previously existing landfills did not include radioactivity limits; instead, these levels were administratively controlled. Typically, if radioactivity was detected above background levels, the waste was classified as low-level waste (LLW), which would be sent off-site for disposal

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

International Nuclear Information System (INIS)

Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

1991-01-01

Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

Science.gov (United States)

Moore, R.H.

1962-10-01

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

Processing of nuclear power plant waste streams containing boric acid

International Nuclear Information System (INIS)

1996-10-01

Boric acid is used in PWR type reactor's primary coolant circuit to control the neutron flux. However, boric acid complicates the control of water chemistry of primary coolant and the liquid radioactive waste produced from NPP. The purpose of this report is to provide member states with up-to-date information and guidelines for the treatment and conditioning of boric acid containing wastes. It contains chapters on: (a) characteristics of waste streams; (b) options for management of boric acid containing waste; (c) treatment/decontamination of boric acid containing waste; (d) concentration and immobilization of boric acid containing waste; (e) recovery and re-use of boric acid; (f) selected industrial processes in various countries; and (g) the influence of economic factors on process selection. 72 refs, 23 figs, 5 tabs

Characterization of PAH matrix with monazite stream containing uranium, gadolinium and iron

Energy Technology Data Exchange (ETDEWEB)

Pal, Sangita, E-mail: sangpal@barc.gov.in; Goswami, D. [Desalination Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400 085 (India); Meena, Sher Singh [Solid State Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400 085 (India)

2016-05-23

Uranium (U) gadolinium (Gd) and iron (Fe) containing alkaline waste simulated effluent (relevant to alkaline effluent of monazite ore) has been treated with a novel amphoteric resin viz, Polyamidehydroxamate (PAH) containing amide and hydroxamic acid groups. The resin has been synthesized in an eco-friendly manner by polymerization nad conversion to functional groups characterized by FT-IR spectra and architectural overview by SEM. Coloration of the loaded matrix and de-coloration after extraction of uranium is the special characteristic of the matrix. Effluent streams have been analyzed by ICP-AES, U loaded PAH has been characterized by FT-IR, EXAFS, Gd and Fe by X-ray energy values of EDXRF at 6.053 KeVand 6.405 KeV respectively. The remarkable change has been observed in Mössbauer spectrum of Fe-loaded PAH samples.

Simultaneous treatment of SO2 containing stack gases and waste water

Science.gov (United States)

Poradek, J. C.; Collins, D. D. (Inventor)

1978-01-01

A process for simultaneously removing sulfur dioxide from stack gases and the like and purifying waste water such as derived from domestic sewage is described. A portion of the gas stream and a portion of the waste water, the latter containing dissolved iron and having an acidic pH, are contacted in a closed loop gas-liquid scrubbing zone to effect absorption of the sulfur dioxide into the waste water. A second portion of the gas stream and a second portion of the waste water are controlled in an open loop gas-liquid scrubbing zone. The second portion of the waste water contains a lesser amount of iron than the first portion of the waste water. Contacting in the openloop scrubbing zone is sufficient to acidify the waste water which is then treated to remove solids originally present.

Recovery or removal of uranium contained in small quantity in waste water by tannic-group adsorbent

Energy Technology Data Exchange (ETDEWEB)

Komoto, Shigetoshi [Power Reactor and Nuclear Fuel Development Corp., Kamisaibara, Okayama (Japan). Ningyo Toge Works

1991-12-01

It was found that tannic compounds have a very strong affinity for uranium and thorium which are nuclear fuel materials, and the new uranium adsorbents composed mainly by tannic-group compounds were made. The solid-state refractory persimmon tannins in those compounds has the most superior capacity for uranium as high as 1.7 g of uranium on 1 g of the adsorbent. The tests adsorbing uranium on the adsorbent are carried out practically by using dam water of Ningyo-toge Works, PNC. Adsorption tests changed the pH or temperature of dam water, elution test, and adsorption-elution repeating test were performed, and it was found that uranium in dam water contained from ppb-level to ppm-level was recovered or removed with very excellent efficiency. (author).

recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories

International Nuclear Information System (INIS)

Othman, S.H.A.

2003-01-01

reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian fuel manufacture pilot plant) gives rise to a variety of wastes in which the uranium content is of considerable importance. the recovery of uranium from concentrated mother liquors produced from ADU (ammonium diuranate ) precipitation, as well as those due to ADU washing is studied in this work. column of poly-trifluoro-monochloro-ethilene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains uranium .impurities are eluted with 6.5 M HCl, and the uranium is eluted with water and the recovery of uranium is better than 94%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model

Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

Energy Technology Data Exchange (ETDEWEB)

Golovich, Elizabeth C.; Wellman, Dawn M.; Serne, R. Jeffrey; Bovaird, Chase C.

2011-09-30

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments.

Decree of the 23 February 2017 in application of the decree nr 2017-231 of the 23 February 2017 in application of the article L. 542-1-2 of the Code of the Environment, and establishing prescriptions of the National Plan of Management of Radioactive Materials and Wastes

International Nuclear Information System (INIS)

Royal, Segolene

2017-01-01

This decree defines legal arrangements about the implementation of the French national plan of management of radioactive materials and wastes (PNGMDR). It contains some general arrangements, issues related to radioactive materials (depleted uranium, uranium from reprocessing, plutonium and spent fuels, thorium-containing materials), and issues related to the long term management of radioactive wastes. These issues notably concern historical storages and the different types of radioactive wastes: very low level radioactive wastes, short lived low and intermediate level wastes, long-lived intermediate level wastes, long-lived high and intermediate level wastes, radioactive wastes requiring specific works such as mercury-containing wastes, organic liquids and oils, tritiated wastes, and so on, and uranium tailings

National Public Information Symposium on Peaceful Uses of Nuclear Energy, NUC Info' 2000. Radioactive Waste Management and Site Restoration in Uranium Industry. Proceedings. Volume 2

International Nuclear Information System (INIS)

Dobos, Ion; Comsa, Olivia

2000-01-01

These proceedings published in two volumes contain materials presented at the National Public Information Symposium on Peaceful Uses of Nuclear Energy, NUC Info' 2000. Radioactive Waste Management and Site Restoration in Uranium Industry - held on 5th September to 8th September 2000 at Baita - Bihor, Romania. The proceedings are structured in 4 sections: 1. Management of radioactive wastes arising from uranium mining, milling and decommissioning; 2. Uranium mine closing down; 3. Environmental restoration of uranium mining and milling sites; 4. Management of radioactive wastes arising from nuclear applications. The contributions in this volume debate the issues of environment restoration at uranium ore mining and management of radioactive wastes resulted from nuclear applications

Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

Cementation of liquid radioactive waste with high content of borate salts

International Nuclear Information System (INIS)

Gorbunova, O.

2015-01-01

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

Vitrification of liquid waste from nuclear power plants

International Nuclear Information System (INIS)

Sheng Jiawei; Choi, Kwansik; Song, Myung-Jae

2001-01-01

Glass is an acceptable waste form to solidify the low-level waste from nuclear power plants (NPPs) because of the simplicity of processing and its unique ability to accept a wide variety of waste streams. Vitrification is being considered to solidify the high-boron-containing liquid waste generated from Korean NPPs. This study dealt with the development of a glass formulation to solidify the liquid waste. Studies were conducted in a borosilicate glass system. Crucible studies have been performed with surrogate waste. Several developed glass frits were evaluated to determine their suitability for vitrifying the liquid waste. The results indicated that the 20 wt% waste oxides loading required could not be obtained using these glass frits. Flyash produced from coal-burning electric power stations, whose major components are SiO 2 and Al 2 O 3 , is a desirable glass network former. Detailed product evaluations including waste loading, homogeneity, chemical durability and viscosity, etc., were carried out on selected formulations using flyash. Up to 30 wt% of the waste oxides was successfully solidified into the flyash after the addition of 5-10 wt% Na 2 O at 1200 deg. C

Decay calculations on medium-level and actinide-containing wastes from the LWR fuel cycle. Pt. 2

International Nuclear Information System (INIS)

Haug, H.O.

1981-12-01

1. The radiotoxicity index as inherent property of the radionuclide inventory was calculated for medium-level and actinide-containing wastes. The calculations were based on the annual limits of intake of the German Radiation Protection Ordinance as well as the new values of annual limits of intake from ICRP-30. The latter imply a higher rating of the toxicity of transuranium nuclides and a lower rating of Sr-90, Tc-99, and Ra-226. Thus, the annual radiotoxicity index is controlled by the transuranics after 10 to 100 years. 2. From the comparison of the radiotoxicity index of conditional and packed wastes with the same volume of uranium ore, it was evaluated that the relative radiotoxicity of the medium-level wastes decreases below the level of pitchblende after less than 100 years and below a 3% uranium ore after less than 2000 of decay. However, based on ICRP-30, the relative radiotoxicity index decreases below the level of pitchblende after 1000 years and decays to the level of the 3% uranium ore at about 10 5 years. 3. The comparison of the radiotoxicity concentration of the total disposal layer with a uranium ore deposit shows that the radiotoxicity concentration based on ICRP-30 of the self-heating wastes placed in single boreholes decays within 2000 years (high level waste within 3000 years) below the level of a uranium ore deposit of 0.2% uranium. The radiotoxicity concentration of the medium-level process waste and the alpha-waste disposed off in disposal chambers decreases to the level of a uranium ore deposit with 0.4 to 6% uranium after about 10 4 years, and 1% after about 10 5 years. (orig./HP) [de

Solid and liquid radioactive waste treatment

International Nuclear Information System (INIS)

Rzyski, B.M.

1989-01-01

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

Separation of uranium by biosorption

International Nuclear Information System (INIS)

Volesky, B.; Tsezos, M.

1983-01-01

This invention relates to metal ion separation processes and more particularly to processes of extraction of specific ions of or containing heavy metals from waste liquids by means of biosorption. The invention is based upon the discovery that the biomass produced as the result of the growth of a certain carefully selected microbial genus, namely species of the genus Rhizopus, such as Rhizopus arihizus, has an outstanding ability for selective adsorption of uranium and thorium ions from aqueous solution or suspension

CHARACTERISATION OF SOLID AND LIQUID PINEAPPLE WASTE

Directory of Open Access Journals (Sweden)

Abdullah Abdullah

2011-07-01

Full Text Available The pineapple waste is contain high concentration of biodegradable organic material and suspended solid. As a result it has a high BOD and extremes of pH conditions. The pineapple wastes juice contains mainly sucrose, glucose, fructose and other nutrients. The characterisation this waste is needed to reduce it by  recycling to get raw material or  for  conversion into useful product of higher value added products such as organic acid, methane , ethanol, SCP and enzyme. Analysis of sugar indicates that liquid waste contains mainly sucrose, glucose and fructose.  The dominant sugar was fructose, glucose and sucrose.  The fructose and glucose levels were similar to each other, with fructose usually slightly higher than glucose. The total sugar and citric acid content were 73.76 and 2.18 g/l. The sugar content in solid waste is glucose and fructose was 8.24 and 12.17 %, no sucrose on this waste

Method of processing radioactive cesium liquid wastes

International Nuclear Information System (INIS)

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

1985-01-01

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

Operation of the D0 uranium liquid-argon calorimeter system

International Nuclear Information System (INIS)

Guida, J.

1992-12-01

The DO calorimeter consists of three separate cryostats containing uranium modules in liquid argon. This odorimeter has transverse segmentation of 0.1 x 0.1 in η x 0 and consists of eight or nine longitudinal readout segments. The coverage in η extends to 4. As a result of the large coverage and fine segmentation there are 50,000 channels of electronics. After a brief description of the electronics, stability and noise aspects will be investigated. Results of the liquid-argon purity studies will be discssed. The backgrounds in the calorimeter due to the Fermilab main ring will also be examined

Decontamination liquid waste processing method

International Nuclear Information System (INIS)

Enda, Masami; Hosaka, Katsumi.

1992-01-01

Liquid wastes after electrolytic reduction are caused to flow through an anionic exchange membrane in a diffusion dialysis step, and liquid wastes and dialyzed water are passed in a countercurrent manner. Since acids in the liquid wastes transfer on the side of the dialyzed water due to the difference of concentration between the liquid wastes and the dialyzed water, acids can be easily recovered from the liquid wastes. If the acid-removed liquid wastes are put to electrodeposition in an electrodepositing step, the electrodepositing reactions between radioactive materials such as Co ion, Mn ion and leached metals such as Fe ions and Cr ions are caused preferentially to hydrogen generation reaction on a metal deposition cathode. Accordingly, metal ions can be easily separated from the liquid wastes. Since the separated liquid wastes are an aqueous solution in which cerium ions as a decontaminant and an acid at low concentration are dissolved, the concentration thereof is controlled by mixing them to acid recovering water after the diffusion dialysis and they can be reused as the decontaminant. (T.M.)

Selective Removal of Uranium from the Washing Solution of Uranium-Contaminated Soil

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

This study examined selective removal methods of uranium from the waste solution by ion exchange resins or solvent extraction methods to reduce amount of the 2{sup nd} waste. Alamine-336, known as an excellent extraction reagent of uranium from the leaching solution of uranium ore, did not remove uranium from the acidic washing solution of soil. Uranyl ions in the acidic waste solution were sorbed on ampholyte resin with a high sorption efficiency, and desorbed from the resin by a washing with 0.5 M Na{sub 2}CO{sub 3} solution at 60 .deg. C. However, the uranium dissolved in the sulfuric acid solution was not sorbed onto the strong anion exchanger resins. A great amount of uranium-contaminated (U-contaminated) soil had been generated from the decommissioning of a uranium conversion plant. Our group has developed a decontamination process with washing and electrokinetic methods to decrease the amount of waste to be disposed of. However, this process generates a large amount of waste solution containing various metal ions.

URANIUM BISMUTHIDE DISPERSION IN MOLTEN METAL

Science.gov (United States)

Teitel, R.J.

1959-10-27

The formation of intermetallic bismuth compounds of thorium or uranium dispersed in a liquid media containing bismuth and lead is described. A bismuthide of uranium dispersed in a liquid metal medium is formed by dissolving uranium in composition of lead and bismuth containing less than 80% lead and lowering the temperature of the composition to a temperature below the point at which the solubility of uranium is exceeded and above the melting point of the composition.

Cerium, uranium, and plutonium behavior in glass-bonded sodalite, a ceramic nuclear waste form

International Nuclear Information System (INIS)

Lewis, M. A.; Lexa, D.; Morss, L. R.; Richmann, M. K.

1999-01-01

Glass-bonded sodalite is being developed as a ceramic waste form (CWF) to immobilize radioactive fission products, actinides, and salt residues from electrometallurgical treatment of spent nuclear reactor fuel. The CWF consists of about 75 mass % sodalite, 25 mass % glass, and small amounts of other phases. This paper presents some results and interpretation of physical measurements to characterize the CWF structure, and dissolution tests to measure the release of matrix components and radionuclides from the waste form. Tests have been carried out with specimens of the CWF that contain rare earths at concentrations similar to those expected in the waste form. Parallel tests have been carried out on specimens that have uranium or plutonium as well as the rare earths at concentrations similar to those expected in the waste forms; in these specimens UCl 3 forms UO 2 and PuCl 3 forms PuO 2 . The normalized releases of rare earths in dissolution tests were found to be much lower than those of matrix elements (B, Si, Al, Na). When there is no uranium in the CWF, the release of cerium is two to ten times lower than the release of the other rare earths. The low release of cerium may be due to its tetravalent state in uranium-free CWF. However, when there is uranium in the CWF, the release of cerium is similar to that of the other rare earths. This trivalent behavior of cerium is attributed to charge transfer or covalent interactions among cerium, uranium, and oxygen in (U,Ce)O 2

Cleaning of spent solvent and method of processing cleaning liquid waste

International Nuclear Information System (INIS)

Ozawa, Masaki; Kawada, Tomio; Tamura, Nobuhiko.

1993-01-01

Spent solvents discharged from a solvent extracting step mainly comprise n-dodecane and TBP and contain nuclear fission products and solvent degradation products. The spent solvents are cleaned by using a sodium chloride free detergent comprising hydrazine oxalate and hydrazine carbonate in a solvent cleaning device. Nitric acid is added to the cleaning liquid wastes containing spent detergents extracted from the solvent cleaning device, to control an acid concentration. The detergent liquid wastes of controlled acid concentration are sent to an electrolysis oxidation bath as electrolytes and electrochemically decomposed in carbonic acid gas, nitrogen gas and hydrogen gas. The decomposed gases are processed as off gases. The decomposed liquid wastes are processed as a waste nitric acid solution. This can provide more effective cleaning. In addition, the spent detergent can be easily decomposed in a room temperature region. Accordingly, the amount of wastes can be decreased. (I.N.)

Program for certification of waste from contained firing facility: Establishment of waste as non-reactive and discussion of potential waste generation problems

International Nuclear Information System (INIS)

Green, L.; Garza, R.; Maienschein, J.; Pruneda, C.

1997-01-01

Debris from explosives testing in a shot tank that contains 4 weight percent or less of explosive is shown to be non-reactive under the specified testing protocol in the Code of Federal Regulations. This debris can then be regarded as a non-hazardous waste on the basis of reactivity, when collected and packaged in a specified manner. If it is contaminated with radioactive components (e.g. depleted uranium), it can therefore be disposed of as radioactive waste or mixed waste, as appropriate (note that debris may contain other materials that render it hazardous, such as beryllium). We also discuss potential waste generation issues in contained firing operations that are applicable to the planned new Contained Firing Facility (CFF). The goal of this program is to develop and document conditions under which shot debris from the planned Contained Firing Facility (CFF) can be handled, shipped, and accepted for waste disposal as non-reactive radioactive or mixed waste. This report fulfills the following requirements as established at the outset of the program: 1. Establish through testing the maximum level of explosive that can be in a waste and still have it certified as non-reactive. 2. Develop the procedure to confirm the acceptability of radioactive-contaminated debris as non-reactive waste at radioactive waste disposal sites. 3. Outline potential disposal protocols for different CFF scenarios (e.g. misfires with scattered explosive)

Licensing of uranium mine and mill waste management systems

International Nuclear Information System (INIS)

Chamney, L.G.

1986-09-01

Systems for the management of wastes arising from uranium mining facilities are subject to regulatory control by the Atomic Energy Control Board (AECB). This paper describes the primary objectives, principles, requirements and guidelines which the AECB uses in the regulation of waste management activities at uranium mining facilities, and provides an understanding of the licensing process used by the AECB

Natural diatomite process for removal of radioactivity from liquid waste

Energy Technology Data Exchange (ETDEWEB)

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

2007-01-15

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

Natural diatomite process for removal of radioactivity from liquid waste

International Nuclear Information System (INIS)

Osmanlioglu, Ahmet Erdal

2007-01-01

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

Natural diatomite process for removal of radioactivity from liquid waste.

Science.gov (United States)

Osmanlioglu, Ahmet Erdal

2007-01-01

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

Improvement for waste water treatment process of a uranium deposite and its effect

International Nuclear Information System (INIS)

Huang Jimao

2013-01-01

Uranium was recovered from alkaline uranium ores by heap leaching and traditional agitation leaching methods at a uranium mine, and the waste water (including waste water produced in hydrometallurgy process and mine drainage) was treated by using chemical precipitation method and chemical precipitation loading method. It was found that the removal rate of uranium by the waste water treatment process was not satisfactory after one year's run. So, the waste water treatment process was improved. After the improvement, removal rate of CO 3 2- ,HCO 3 - , U and Ra was enhanced and the treated waste water reached the standard of discharge. (author)

Process for recovering a uranium containing concentrate and purified phosphoric acid from a wet process phosphoric acid containing uranium

International Nuclear Information System (INIS)

Weterings, C.A.M.; Janssen, J.A.

1985-01-01

A process is claimed for recovering from a wet process phosphoric acid which contains uranium, a uranium containing concentrate and a purified phosphoric acid. The wet process phosphoric acid is treated with a precipitant in the presence of a reducing agent and an aliphatic ketone

Process for recovering a uranium containing concentrate and purified phosphoric acid from a wet process phosphoric acid containing uranium

Energy Technology Data Exchange (ETDEWEB)

Weterings, C.A.M.; Janssen, J.A.

1985-04-30

A process is claimed for recovering from a wet process phosphoric acid which contains uranium, a uranium containing concentrate and a purified phosphoric acid. The wet process phosphoric acid is treated with a precipitant in the presence of a reducing agent and an aliphatic ketone.

Radioactive safety analysis and assessment of waste rock pile site in uranium tailings

International Nuclear Information System (INIS)

Liu Changrong; Liu Zehua; Wang Zhiyong; Zhou Xinghuo

2007-01-01

Based on theoretical calculation and in-situ test results, distribution and emissions of radioactive nuclides of uranium tailings impoundment and waste rock pile sites are analyzed in this paper. It is pointed out that 222 Rn is the main nuclide of uranium tailings impoundment and waste rock pile site. Also 222 Rn is the main source term of public dose. 222 Rn concentrations in the atmospheric environment around and individual dose to Rn gradually decrease with increasing distances to uranium tailings impoundment and waste rock pile site. Based on in-situ tests on five uranium tailings impoundment and waste rock pile sites, a decisive method and safety protection distance are presented, which can be used to guide the validation and design of radioactive safety protection in uranium tailings impoundment and waste rock pile sites. (authors)

Processing method and device for radioactive liquid waste

International Nuclear Information System (INIS)

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

1997-01-01

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

Comparative analysis of bacteria in uranium mining wastes

International Nuclear Information System (INIS)

Tzvetkova, T.; Flemming, K.; Selenska-Pobell, S.

2002-01-01

Compositional analysis of predominant bacterial groups in three different kinds of uranium wastes gives indications for different biogeological processes running at the studied sites which seems to be influenced by the anthropological activities involved in the production of uranium. (orig.)

Physicochemical basics for production of uranium concentrate from wastes of hydrometallurgical plants and technical waters

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Khojiyon, M.; Mirsaidov, I.U.; Nazarov, K.M.; Barotov, B.B.

2012-01-01

Physicochemical and technological basics for reprocessing of uranium industry wastes of Northern Tajikistan shows that the most perspective for reprocessing is Chkalovsk tailing's wastes. Engineer and geological condition and content of radionuclides in wastes are investigated. It is determined that considered wastes by radioactivity are low-active and they can be reprocessed with the purpose of U 3 O 8 production. Grinding, crumbling, thickening and etc. operations are decreased during the wastes reprocessing process. Uranium output is more than 90%. Optimal parameters of products extraction from uranium mining industry wastes are found. Characteristics of mine and technical waters of uranium industry wastes are studied. Characteristics of mine and technical waters of Kiik-Tal and Istiklol city (former Taboshar) showed the expediency of uranium oxide extraction from them. The reasons for non-additional recovery extraction from dumps of State Enterprise 'Vostokredmet' by classical methods of uranium leaching are studied. Kinetics of sulfuric leaching of residues from anthropogenic deposit of Map 1-9 (Chkalovsk city) is investigated. Carried out investigations are revealing the flow mechanism process of residues' sulfuric leaching and enable selection of radiation regime of U 3 O 8 production. Kinetics of sorption process of uranium extraction from mine and technical waters of uranium industry wastes is studied. High sorption properties of apricot's shell comparing to other sorbents are revealed. Basic process flow diagram for reprocessing of uranium tailing wastes is developed as well as diagram for uranium extraction from mine and technical waters from uranium industry wastes which consists of the following stages: acidification, sorption, burning, leaching, sedimentation, filtration, drying.

Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification.

Science.gov (United States)

de Blois, Erik; Chan, Ho Sze; Roy, Kamalika; Krenning, Eric P; Breeman, Wouter A P

PET with 68 Ga from the TiO 2 - or SnO 2 - based 68 Ge/ 68 Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity ( 68 Ge vs. 68 Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of 68 Ge activity is produced by eluting the 68 Ge/ 68 Ga generators and residues from PET chemistry. Since clearance level of 68 Ge activity in waste may not exceed 10 Bq/g, as stated by European Directive 96/29/EURATOM, our purpose was to reduce 68 Ge activity in solution from >10 kBq/g to <10 Bq/g; which implies the solution can be discarded as regular waste. Most efficient method to reduce the 68 Ge activity is by sorption of TiO 2 or Fe 2 O 3 and subsequent centrifugation. The required 10 Bq per mL level of 68 Ge activity in waste was reached by Fe 2 O 3 logarithmically, whereas with TiO 2 asymptotically. The procedure with Fe 2 O 3 eliminates ≥90% of the 68 Ge activity per treatment. Eventually, to simplify the processing a recirculation system was used to investigate 68 Ge activity sorption on TiO 2 , Fe 2 O 3 or Zeolite. Zeolite was introduced for its high sorption at low pH, therefore 68 Ge activity containing waste could directly be used without further interventions. 68 Ge activity containing liquid waste at different HCl concentrations (0.05-1.0 M HCl), was recirculated at 1 mL/min. With Zeolite in the recirculation system, 68 Ge activity showed highest sorption.

Recent advances in liquid membranes and their applications in nuclear waste processing: an overview

Energy Technology Data Exchange (ETDEWEB)

Shukla, J P; Iyer, R H [Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Membrane extraction, combining the processes of extraction, scrubbing and stripping in a single step, demonstrates the inherent capability of solvent extraction under non-equilibrium conditions. Permeant transport across various liquid membrane (LM) configurations, viz. bulk liquid, emulsion liquid and supported liquid membranes has great potential for applications in the nuclear field particularly in the decontamination of low and medium level radioactive wastes. Potential practical applications of such membranes have also been envisaged in the recovery of metals from hydrometallurgical leach solutions and in plutonium and americium removal from nitric acid waste streams generated by plutonium recovery operations in the PUREX process. Studies carried out have established that minor actinides like uranium, plutonium and americium from process effluents can easily be transported across polymeric and liquid type membranes through the use of specific ionophores dissolved in an appropriate liquid membrane phase. The possibility of the membrane extraction of fission palladium from acidic wastes has also been demonstrated by the use of some soft bases. An overview of these results and also some of the recent radiochemical applications of energy - efficient LM processes including directions for future research are outlined in this paper. (author). 19 refs., 1 fig., 2 tabs.

The nondestructive assay of 55-gallon drums containing uranium and transuranic waste using passive-active shufflers

International Nuclear Information System (INIS)

Rinard, P.M.; Adams, E.L.; Menlove, H.O.; Sprinkle, J.K. Jr.

1992-11-01

This study has been completed to characterize and improve the performance of passive-active neutron (PAN) shufflers in assaying 55gal. drums of nuclear facility waste for uranium and transuranic elements. Over 1700 active measurements and 800 passive measurements were made using 28 different matrices. Some of the matrices had homogeneous distributions of known amounts of moderating and absorbing materials, whereas others were less well characterized. Some of the well-characterized matrices simulate facility waste better than the others,especially matrices of paper, iron, polyethylene in nine different densities (with and without neutron poisons), alumina trap material, and concrete blocks

Obtain of uranium concentrates from fertil liquids

International Nuclear Information System (INIS)

Narvaez Castillo, W.A.

1992-01-01

This research tried to encounter the form to remove uranium from the rock in the best way, for that it was used different process like leaching, extraction, concentration and precipitation. To leach the mineral was chosen basic leaching, using a mixture of carbonate-sodium bicarbonate, this method is more adequated for the basic nature of the mineral. In extraction was used specific uranium ionic interchanges, so was chosen a tertiary amine like Alamina 336. The concentration phase is intimately binding with the extraction by ionic interchange, for the capability of resine's extraction to obtain concentrated liquids. When the liquids were obtained with high concentration of uranium in the same time were purified and then were precipitated, for that we employed a precipitant agent like: Sodium hydroxide, Amonium hydroxide, Magnesium hydroxide, Hydrogen peroxide and phosphates. With all concentrates we obtain the YELLOW CAKE

Method of processing concentrated liquid waste in nuclear power plant

International Nuclear Information System (INIS)

Hasegawa, Kazuyuki; Kitsukawa, Ryozo; Ohashi, Satoru.

1988-01-01

Purpose: To reduce the oxidizable material in the concentrated liquid wastes discharged from nuclear power plants. Constitution: Nitrate bacteria are added to liquid wastes in a storage tank for temporarily storing concentrated liquid wastes or relevant facilities thereof. That is, nitrites as the oxidizable material contained in the concentrated liquid wastes are converted into nitrate non-deleterious to solidification by utilizing biological reaction of nitrate bacteria. For making the conversion more effectively, required time for the biological reaction of the nitrate bacteria is maintained from the injection of nitrate bacteria to solidification, thereby providing advantageous conditions for the propagation of the nitrate bacteria. In this way, there is no problem for the increase of the volume of the powdery wastes formed by the addition of inhibitor for the effect of oxidizable material. Further, heating upon solidification which is indispensable so far is no more necessary to simplify the facility and the operation. Furthermore, the solidification inhibiting material can be reduced stably and reliably under the same operation conditions even if the composition of the liquid wastes is charged or varied. (Kamimura, M.)

Liquid effluent retention facility dangerous waste permit application

International Nuclear Information System (INIS)

1991-06-01

This appendix to the Liquid Effluent Retention Facility Dangerous Waste Permit Application contains pumps, piping, leak detection systems, geomembranes, leachate collection systems, earthworks and floating cover systems

Selective separation of uranium from nuclear waste solution by bis (2,4,4-trimethylpentyl phosphinic) acid in ionic liquid and molecular diluents: a comparative study

International Nuclear Information System (INIS)

Singh, Manpreet; Sengupta, Arijit; Murali, M.S.; Adya, V.C.; Kadam, R.M.

2016-01-01

Room temperature ionic liquid has been world-wide considered as the potential 'green' alternatives to the molecular diluents. A comparative study was carried out for studying selective separation of uranium from radioactive waste solution using Bis(2,4,4-trimethylpentyl phosphinic) acid in molecular diluent (xylene) and ionic liquid (C 8 mimNTf 2 ). For ionic liquid based system, the extraction kinetics was found to be slower compared to the molecular diluents. This was attributed to the higher viscosity of ionic liquid. In ionic liquid the extraction occurs with the predominance of 'ion exchange' mechanism through (UO 2 (NO 3 ). 2L) + species, while for xylene based system 'solvation' mechanism predominates at higher feed acidity. The extraction process in ionic liquid was found to be thermodynamically more favoured than in xylene. The nature of the extracted species was found to be different in ionic liquid and xylene as obtained from difference in luminescence emission profiles and lifetime of the extracted complex. Ionic liquid based system was found to be radiolytically more stable than the molecular diluents based solvent system. Na 2 CO 3 solution was found to back extract the uranyl ion almost quantitatively (99.9 %) from the loaded organic phase but overall stripping from ionic liquid phase is comparatively poorer than that of xylene phase. The processing of Simulated High Level Waste (SHLW) of Pressurized Heavy Water Reactor (PHWR) or Research Reactor (RR) origin revealed that bis(2,4,4-trimethylpentyl phosphinic) acid can effectively be used for the preferential extraction of U with better selectivity for ionic liquid phase. But the ion exchange mechanism is one of the disadvantages for its plant scale application. (author)

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

Uranium material removing and recovering device

International Nuclear Information System (INIS)

Takita, Shin-ichi.

1997-01-01

A uranium material removing and recovering device for use in removing surplus uranium heavy metal (UO 2 ) generated in a uranium handling facility comprises a uranium material removing device and a uranium material recovering device. The uranium material removing device comprises an adsorbing portion filled with a uranium adsorbent, a control portion for controlling the uranium adsorbent of the uranium adsorbing portion by a controlling agent, a uranium adsorbing device connected thereto and a jetting device for jetting the adsorbing liquid to equipments deposited with uranium. The recovering device comprises a recovering apparatus for recovering uranium materials deposited with the adsorbent liquid removed by the jetting device and a recovering tank for storing the recovered uranium materials. The device of the present invention can remove surplus uranium simply and safely, mitigate body's load upon removing and recovering operations, facilitate the processing for the exchange of the adsorbent and reduces the radioactive wastes. (T.M.)

Water quality for liquid wastes

International Nuclear Information System (INIS)

Mizuniwa, Fumio; Maekoya, Chiaki; Iwasaki, Hitoshi; Yano, Hiroaki; Watahiki, Kazuo.

1985-01-01

Purpose: To facilitate the automation of the operation for a liquid wastes processing system by enabling continuous analysis for the main ingredients in the liquid wastes accurately and rapidly. Constitution: The water quality monitor comprises a sampling pipeway system for taking out sample water for the analysis of liquid wastes from a pipeway introducing liquid wastes to the liquid wastes concentrator, a filter for removing suspended matters in the sample water and absorption photometer as a water quality analyzer. A portion of the liquid wastes is passed through the suspended matter filter by a feedpump. In this case, sulfate ions and chloride ions in the sample are retained in the upper portion of a separation color and, subsequently, the respective ingredients are separated and leached out by eluting solution. Since the leached out ingredients form ferric ions and yellow complexes respectively, their concentrations can be detected by the spectrum photometer. Accordingly, concentration for the sodium sulfate and sodium chloride in the liquid wastes can be analyzed rapidly, accurately and repeatedly by which the water quality can be determined rapidly and accurately. (Yoshino, Y.)

Physicochemical basics for production of uranium concentrate from wastes of hydrometallurgical plants and technical waters

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Khojiyon, M.; Mirsaidov, I.U.; Nazarov, K.M.; Barotov, B.B.

2012-01-01

Physicochemical and technological basics for reprocessing of uranium industry wastes of Northern Tajikistan shows that the most perspective for reprocessing is Chkalovsk tailing's wastes. Engineer and geological condition and content of radionuclides in wastes are investigated. It is determined that considered wastes by radioactivity are low-active and they can be reprocessed with the purpose of U 3 O 8 production. Grinding, crumbling, thickening and etc. operations are decreased during the wastes reprocessing process. Uranium output is more than 90%. Optimal parameters of products extraction from uranium mining industry wastes are found. Characteristics of mine and technical waters of uranium industry wastes are studied. Characteristics of mine and technical waters of Kiik-Tal and Istiklol city (former Taboshar) showed the expediency of uranium oxide extraction from them. The reasons for non-additional recovery extraction from dumps of State Enterprise 'Vostokredmet' by classical methods of uranium leaching are studied. Kinetics of sulfuric leaching of residues from anthropogenic deposit of Map 1-9 (Chkalovsk city) is investigated. Carried out investigations are revealing the flow mechanism process of residues' sulfuric leaching and enable selection of radiation regime of U 3 O 8 production. Kinetics of sorption process of uranium extraction from mine and technical waters of uranium industry wastes is studied. High sorption properties of apricot's shell comparing to other sorbents are revealed. Basic process flow diagram for reprocessing of uranium tailing wastes is developed as well as diagram for uranium extraction from mine and technical waters from uranium industry wastes which consists of the following stages: acidification, sorption, burning, leaching, sedimentation, filtration, drying.

REDUKSI AKTIVITAS URANIUM DALAM LIMBAH RADIOAKTIF CAIR MENGGUNAKAN PROSES ELEKTROKOAGULASI

Directory of Open Access Journals (Sweden)

Prayitno Prayitno

2017-01-01

Full Text Available ABSTRAK REDUKSI AKTIVITAS URANIUM DALAM LIMBAH RADIOAKTIF CAIR MENGGUNAKAN PROSES ELEKTROKOAGULASI. Limbah yang dihasilkan dari proses pengembangan bahan industri bersifat radioaktif yang mengandung uranium yang dapat menimbulkan dampak negatif pada manusia dan lingkungan. Pengolahan limbah radioaktif pada saat ini masih banyak menggunakan bahan-bahan kimia.Penambahanbahan kimiauntuk mereduksi bahan pencemar dinilai kurang efisien karena kurang ramah lingkungan, memerlukan waktu yang lama, dan biaya yang mahal. Untuk itu akan diterapkan metode proses elektrokoagulasi untuk menurunkan aktivitas uranium dari larutan limbah cair. Tujuan penelitian ini adalah untuk mengetahui efisiensi penurunan aktivitas uranium dalam limbah radioaktif cair yang dihasilkan pada proses elektrokoagulasi dengan variasi tegangan, waktu tinggal, jarak elektroda dan pH inlet limbah. Limbah simulasi yang digunakan memiliki kadar kontaminan uranium sebesar 500 mg/L. Percobaan ini dilakukan dengan metode batch dengan elektroda aluminium. Hasil penelitian diperoleh parameter optimal pada tegangan 12,50 V, jarak 1 cm, pH 7, dan waktu proses selama 60 menit diperoleh efisiensi penurunan limbah uranium sebesar 97,20 %. Kata Kunci:elektrokoagulasi, reduksi limbah uranium, tegangan, aluminium. ABSTRACT REDUCTION OF URANIUM ACTIVITIES IN LIQUID WASTE RADIOACTIVE BY USING OF ELCTROCOAGULATION PROCESS.  Waste generated from the process of the industrial material development one of which waste containing uranium radioactive, can have negative impact on humans and the environment. In  the present time, chemicals are still mostly use in radioactive waste treatment. To reduce pollutants with the use of chemicals is less efficient, because less environmentally friendly, take long time and costly. Therefore, a system of electrocoagulation process will be applied to decrease the activity of uranium from waste solution. The purpose of this study is to determine the efficiency of uranium

Mixed Waste Integrated Program: A technology assessment for mercury-containing mixed wastes

International Nuclear Information System (INIS)

Perona, J.J.; Brown, C.H.

1993-03-01

The treatment of mixed wastes must meet US Environmental Protection Agency (EPA) standards for chemically hazardous species and also must provide adequate control of the radioactive species. The US Department of Energy (DOE) Office of Technology Development established the Mixed Waste Integrated Program (MWIP) to develop mixed-waste treatment technology in support of the Mixed Low-Level Waste Program. Many DOE mixed-waste streams contain mercury. This report is an assessment of current state-of-the-art technologies for mercury separations from solids, liquids, and gases. A total of 19 technologies were assessed. This project is funded through the Chemical-Physical Technology Support Group of the MWIP

Nuclear waste viewed in a new light; a synchrotron study of uranium encapsulated in grout.

Science.gov (United States)

Stitt, C A; Hart, M; Harker, N J; Hallam, K R; MacFarlane, J; Banos, A; Paraskevoulakos, C; Butcher, E; Padovani, C; Scott, T B

2015-03-21

How do you characterise the contents of a sealed nuclear waste package without breaking it open? This question is important when the contained corrosion products are potentially reactive with air and radioactive. Synchrotron X-rays have been used to perform micro-scale in-situ observation and characterisation of uranium encapsulated in grout; a simulation for a typical intermediate level waste storage packet. X-ray tomography and X-ray powder diffraction generated both qualitative and quantitative data from a grout-encapsulated uranium sample before, and after, deliberately constrained H2 corrosion. Tomographic reconstructions provided a means of assessing the extent, rates and character of the corrosion reactions by comparing the relative densities between the materials and the volume of reaction products. The oxidation of uranium in grout was found to follow the anoxic U+H2O oxidation regime, and the pore network within the grout was observed to influence the growth of uranium hydride sites across the metal surface. Powder diffraction analysis identified the corrosion products as UO2 and UH3, and permitted measurement of corrosion-induced strain. Together, X-ray tomography and diffraction provide means of accurately determining the types and extent of uranium corrosion occurring, thereby offering a future tool for isolating and studying the reactions occurring in real full-scale waste package systems. Copyright © 2014 Elsevier B.V. All rights reserved.

Risk assessment and quality improvement of liquid waste management in Taiwan University chemical laboratories.

Science.gov (United States)

Ho, Chao-Chung; Chen, Ming-Shu

2018-01-01

The policy of establishing new universities across Taiwan has led to an increase in the number of universities, and many schools have constructed new laboratories to meet students' academic needs. In recent years, there has been an increase in the number of laboratory accidents from the liquid waste in universities. Therefore, how to build a safety system for laboratory liquid waste disposal has become an important issue in the environmental protection, safety, and hygiene of all universities. This study identifies the risk factors of liquid waste disposal and presents an agenda for practices to laboratory managers. An expert questionnaire is adopted to probe into the risk priority procedures of liquid waste disposal; then, the fuzzy theory-based FMEA method and the traditional FMEA method are employed to analyze and improve the procedures for liquid waste disposal. According to the research results, the fuzzy FMEA method is the most effective, and the top 10 potential disabling factors are prioritized for improvement according to the risk priority number (RNP), including "Unclear classification", "Gathering liquid waste without a funnel or a drain pan", "Lack of a clearance and transport contract", "Liquid waste spill during delivery", "Spill over", "Decentralized storage", "Calculating weight in the wrong way", "Compatibility between the container material and the liquid waste", "Lack of dumping and disposal tools", and "Lack of a clear labels for liquid waste containers". After tracking improvements, the overall improvement rate rose to 60.2%. Copyright © 2017 Elsevier Ltd. All rights reserved.

Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

2011-06-08

Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate’s beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ~60°C, 80°C, and 95°C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

International Nuclear Information System (INIS)

Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

2011-01-01

Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate's beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ∼60 C, 80 C, and 95 C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

Groundwater geochemistry near the storage sites of low-level radioactive waste: Implications for uranium migration

Energy Technology Data Exchange (ETDEWEB)

Gaskova, Olga L.; Boguslavsky, Anatoly E. [Institute of Geology and Mineralogy SB RAS, Ac. Koptyug prosp. 3, Novosibirsk 630090 (Russian Federation)

2013-07-01

This paper presents results of detailed sampling of groundwater and surface water near the storage sites of radioactive waste from the Electrochemical Plant ECP (Zelenogorsk, Krasnoyarsk region, Russia) and the Angarsk Electrolysis Chemical Complex AEC (Angarsk, Irkutsk region, Russia), both of which have produced enriched uranium since 1960's. The liquid (LRW) and solid (SRW) radioactive wastes belong to the category of low-level activity waste. The main result is that the uranium is below the recommended MPC for drinking waters in all types of groundwater around the sludge of ECP and AEC. But alkaline nitrate solutions have been penetrating and spreading into the aquifers under the LRW sludge pits. According to our calculations, redox conditions in the groundwater influenced by discharge are controlled by the couple NO{sub 3}{sup -}/NO{sub 2}{sup -} that facilitates U(VI) migration. The groundwater under SRW repositories is distinguished by its low mineralization and neutral pH. Co-contaminants, such as Mo, V, and Zr may serve as markers of techno-genous contamination in storage sites of the LRW sludge. (authors)

Determination of Uranium in Aqueous and Organic Medium From Product and Waste Processes by Potentiometric Titration Using Modified Davies Gray Method

International Nuclear Information System (INIS)

Putro, K.P; Suripto, A

1998-01-01

Determination of uranium in aqueous and organic solution generated from nuclear fuels production and liquid radioactive waste at Fuel Element Production Installation for Research Reactor, by modified Davies-Gray method using phosphoric acid as medium and vanadium as catalyst has been carried out. The performed at different concentration of phosphoric acid, vanadium and the effect of impurities, as Al, Fe, Si, Cl and F in sample are measurement. Determination of uranium in organic solvent are the sample volume, agitation time and the optimum concentration of uranium to measurement. It was observed that, the optimum conditions for uranium analysis were : 5 -400 mg uranium in 3.2 M phosphoric acid medium containing 220 mg/l vanadium as catalyst. The impurities of Al ≤ 40.5 μg/ml, Fe ≤ 67.6 μg/ml, Si ≤ 20.3 μg/ml, Cl ≤ 135.1 μg/ml and F 13.5 μg/ml have not effect, but the concentration of F ≥ 40.5 μg/ml have effect in analysis result. The uranium content detectable in organic medium has been found between 0.01 to 0.10 g/l and the reproducibility range between 0.09 to 0.15 as well as the sample volume should be in the range of 5 and 10 ml by the agitation time for 4 minute

Standard test method for the analysis of refrigerant 114, plus other carbon-containing and fluorine-containing compounds in uranium hexafluoride via fourier-transform infrared (FTIR) spectroscopy

CERN Document Server

American Society for Testing and Materials. Philadelphia

2004-01-01

1.1 This test method covers determining the concentrations of refrigerant-114, other carbon-containing and fluorine-containing compounds, hydrocarbons, and partially or completely substituted halohydrocarbons that may be impurities in uranium hexafluoride. The two options are outlined for this test method. They are designated as Part A and Part B. 1.1.1 To provide instructions for performing Fourier-Transform Infrared (FTIR) spectroscopic analysis for the possible presence of Refrigerant-114 impurity in a gaseous sample of uranium hexafluoride, collected in a "2S" container or equivalent at room temperature. The all gas procedure applies to the analysis of possible Refrigerant-114 impurity in uranium hexafluoride, and to the gas manifold system used for FTIR applications. The pressure and temperatures must be controlled to maintain a gaseous sample. The concentration units are in mole percent. This is Part A. 1.2 Part B involves a high pressure liquid sample of uranium hexafluoride. This method can be appli...

Lining materials for waste disposal containment and waste storage facilities. (Latest citations from the NTIS bibliographic database). Published Search

International Nuclear Information System (INIS)

1993-11-01

The bibliography contains citations concerning the design characteristics, performance, and materials used to make liners for the waste disposal and storage industry. Liners made of concrete, polymeric materials, compacted clays, asphalt, and in-situ glass are discussed. The use of these liners to contain municipal wastes, hazardous waste liquids, and both low-level and high-level radioactive wastes is presented. Liner permeability, transport, stability, construction, and design are studied. Laboratory field measurements for specific wastes are included. (Contains a minimum of 213 citations and includes a subject term index and title list.)

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

International Nuclear Information System (INIS)

Rybalchenko, A.

1998-01-01

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

Characterization and storage of liquid wastes containing 125Iodine in the laboratory for production of brachytherapy sources - IPEN

International Nuclear Information System (INIS)

Carvalho, Vitória S.; Souza, Daiane C.B. de; Barbosa, Nayane K.O.; Rodrigues, Bruna T.; Nogueira, Beatriz R.; Costa, Osvaldo L. da; Zeituni, Carlos A.; Vicente, Roberto; Rostelato, Maria E.C.M.

2017-01-01

Radioactive sources of Iodine-125 for medical applications have been developed at the Institute for Energy and Nuclear Research (IPEN) to meet the growing demand for medical applications such as brachytherapy. A dedicated laboratory is already being implemented at IPEN. Part of the processes involved in the production of sealed sources generate radioactive wastes that despite the short half-life (<100 days) have radioactive activity above the levels of exemption established by the Brazilian National Nuclear Energy Commission. Therefore, these wastes should receive appropriate treatment and storage until they reach the levels of release into the environment. This work aims to determine the volumes of the liquid wastes generated during the production stages of the sources, as well as to propose a temporary storage system for such wastes. The applied methodology consisted in determining the volumes of wastes generated in each production cell according to the manufacturing steps. After that, activities and activity concentrations were calculated for each container used for temporary storage inside the production laboratory. The total volume stored for one year in the temporary storage, as well as the rate of entry and exit of the liquid wastes were calculated according to the source production demand and the decay time of the radionuclide, respectively. The main results showed that the time required to reach sanitary sewage disposal values is within the period of operation of the facility. The total volume generated is also within the facility's temporary storage capacity

Improved ionic model of liquid uranium dioxide

NARCIS (Netherlands)

Gryaznov, [No Value; Iosilevski, [No Value; Yakub, E; Fortov, [No Value; Hyland, GJ; Ronchi, C

The paper presents a model for liquid uranium dioxide, obtained by improving a simplified ionic model, previously adopted to describe the equation of state of this substance [1]. A "chemical picture" is used for liquid UO2 of stoichiometric and non-stoichiometric composition. Several ionic species

Process equipment waste and process waste liquid collection systems

International Nuclear Information System (INIS)

1990-06-01

The US DOE has prepared an environmental assessment for construction related to the Process Equipment Waste (PEW) and Process Waste Liquid (PWL) Collection System Tasks at the Idaho Chemical Processing Plant. This report describes and evaluates the environmental impacts of the proposed action (and alternatives). The purpose of the proposed action would be to ensure that the PEW and PWL collection systems, a series of enclosed process hazardous waste, and radioactive waste lines and associated equipment, would be brought into compliance with applicable State and Federal hazardous waste regulations. This would be accomplished primarily by rerouting the lines to stay within the buildings where the lined floors of the cells and corridors would provide secondary containment. Leak detection would be provided via instrumented collection sumps locate din the cells and corridors. Hazardous waste transfer lines that are routed outside buildings will be constructed using pipe-in-pipe techniques with leak detection instrumentation in the interstitial area. The need for the proposed action was identified when a DOE-sponsored Resource Conservation and Recovery Act (RCRA) compliance assessment of the ICPP facilities found that singly-contained waste lines ran buried in the soil under some of the original facilities. These lines carried wastes with a pH of less than 2.0, which were hazardous waste according to the RCRA standards. 20 refs., 7 figs., 1 tab

Simple and Rapid Dual-Dispersive Liquid-Liquid Microextraction as an Innovative Extraction Method for Uranium in Real Water Samples Prior to the Determination of Uranium by a Spectrophotometric Technique.

Science.gov (United States)

Khan, Naeemullah; Tuzen, Mustafa; Kazi, Tasneem Gul

2017-11-01

An innovative, rapid, and simple dual-dispersive liquid-liquid microextraction (DDLL-ME) approach was used to extract uranium from real samples for the first time. The main objective of this study was to disperse extraction solvent by using an air-agitated syringe system to overcome matrix effects and avoid dispersion of hazardous dispersive organic solvents by using heat. The DDLL-ME method consisted of two dispersive liquid-liquid extraction steps with chloroform as the extracting solvent. Uranium formed complexes with 4-(2-thiazolylazo) resorcinol in the aqueous phase and was extracted in extracting solvent (chloroform) after the first dispersive liquid-liquid process. Uranium was then back-extracted in the acidic aqueous phase in a second dispersive liquid-liquid process. Finally, uranium was determined by a spectrophotometric detection technique. The variables that played a key role in the proposed method were studied and optimized. The LOD and sensitivity enhancement factor for uranium were found to be 0.60 µg/L and 45, respectively, under optimized conditions. Calibration graphs were found to be linear in the range of 5.0-600 µg/L. The RSD was 2.5%. Reliability of the proposed method was verified by analyzing certified reference material TM-28.3.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Salt Removal from the Uranium Deposits of Electrorefiner

Energy Technology Data Exchange (ETDEWEB)

Kwon, S. W.; Park, K. M.; Lee, S. J.; Park, S. B.; Cho, C. H.; Choi, S. Y.; Lee, H. S.; Kim, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps. The deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. In the liquid cathode, cadmium metal should be removed to recover actinide product. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, the solid-liquid separation was proposed prior to distillation of salt and a feasibility of the separation of the liquid salt by a metallic wire mesh (sieve) was tested for the reduction of the burden of the following vacuum distillation process

Salt Removal from the Uranium Deposits of Electrorefiner

International Nuclear Information System (INIS)

Kwon, S. W.; Park, K. M.; Lee, S. J.; Park, S. B.; Cho, C. H.; Choi, S. Y.; Lee, H. S.; Kim, J. G.

2010-01-01

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps. The deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. In the liquid cathode, cadmium metal should be removed to recover actinide product. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, the solid-liquid separation was proposed prior to distillation of salt and a feasibility of the separation of the liquid salt by a metallic wire mesh (sieve) was tested for the reduction of the burden of the following vacuum distillation process

Introduction - Physicochemical and technological aspects of processing of uranium industry wastes in Tajikistan

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

2011-01-01

The uranium deposits of Tajikistan played an immensely significant role in the practical solution of a radioactive raw materials problem which appeared during the post-World War II years in the USSR. The pioneer in this field became complex №6 (currently known as 'Vostokredmet'). The first soviet uranium was produced from the ores extracted from the republic's deposits. For 50 years (1945-1995 y.) , uranium bearing raw materials from all over the former USSR were delivered to Tajikistan, and uranium oxide was produced, which was later delivered back to Russia for further production of enriched uranium. The total volume of uranium produced in Tajikistan plants was approximately 100 thousands tons. In Soghd region, during that period, more than 55 million tons of uranium waste was accumulated. The total activity of the waste, according to different calculations, is approximately 240-285 TBq. The total amount of waste in dumps and tailings piles is estimated to be more than 170 million tons, most of which are located in the neighborhoods of hydrometallurgical plants and heap leaching locations. Uranium industry wastes in Northern Tajikistan have become attractive for different investors and commercial companies, from secondary reprocessing of mines and tailings' point of view, since the uranium price is increasing. In this regard, research on developing uranium extraction methods from wastes is broadening. The study of the possibility and economic reasonability of reprocessing former year's dumps requires comprehensive examination, and relates not only to uranium extraction but to safe extraction of dumps from tailings as well.

Introduction. Physicochemical aspects of uranium concentrates obtaining from the wastes and raw materials

International Nuclear Information System (INIS)

Mirsaidov, I.U.

2014-01-01

The uranium deposits of Tajikistan played an immensely significant role in the practical solution of a radioactive raw materials problem which appeared during the post-World War II years in the USSR. The pioneer in this field became complex №6 (currently known as 'Vostokredmet'). The first soviet uranium was produced from the ores extracted from the republic's deposits. For 50 years (1945-1995 y.), uranium bearing raw materials from all over the former USSR were delivered to Tajikistan, and uranium oxide was produced, which was later delivered back to Russia for further production of enriched uranium. The total volume of uranium produced in Tajikistan plants was approximately 100 thousands tons. In Sughd region, during that period, more than 55 million tons of uranium waste was accumulated. The total activity of the waste, according to different calculations, is approximately 240-285 TBq. The total amount of waste in dumps and tailings piles is estimated to be more than 170 million tons, most of which are located in the neighborhoods of hydrometallurgical plants and heap leaching locations. Uranium industry wastes in Northern Tajikistan have become attractive for different investors and commercial companies, from secondary reprocessing of mines and tailings' point of view, since the uranium price is increasing. In this regard, research on developing uranium extraction methods from wastes is broadening. The study of the possibility and economic reasonability of reprocessing former year's dumps requires comprehensive examination, and relates not only to uranium extraction but to safe extraction of dumps from tailings as well.

Mobilisation and immobilisation of uranium-DOC-species in three waste dumps in the district of Schlema/Alberoda

International Nuclear Information System (INIS)

Kupsch, H.

1998-01-01

The modelling of geochemical processes needs the detailed and comprehensive knowledge of all chemical interactions which are able to exist in the flow path of waste dumps, soils and aquifer. This includes the adsorption, displacement and transport of heavy metal species of fulvic and humic acids which represent the main amount of DOC in the liquid / solid system of flow path. Comparative measurements of DOC concentrations in the input and output flow at the three waste dumps in the district Schlema / Alberoda have indicated that DOC is produced and / or supplied within of waste dumps. The speciation of heavy metal compounds of fulvic and humic acids was realized by means of two methods (sequential chromatographic analysis SCA; ions focussed electrophoresis IFE). Between 5 and 20% of uranium in the flow path exist in the form of fulvic and humic acids species in which the bond of uranium is permanently. Their distribution coefficients are strongly correlated with the pH values of the several geochemical systems. (orig.) [de

Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used.

Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

International Nuclear Information System (INIS)

Kim, Seung Soo; Han, G. S; Park, U. K; Kim, G. N.; Moon, J. K.

2014-01-01

A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used

Method for the disposal of radioactive waste liquids

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Kamiya, K; Kuriyama, O

1976-03-19

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

Development of high integrity containers for rad-waste treatment

Energy Technology Data Exchange (ETDEWEB)

Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1995-12-31

Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

Development of high integrity containers for rad-waste treatment

Energy Technology Data Exchange (ETDEWEB)

Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1996-12-31

Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

Foreword - Physicochemical and technological aspects of processing of uranium industry wastes in Tajikistan

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

2011-01-01

During recent years, the attention of many researchers has turned to decontamination of the territories where radioactive substance extraction took place in the past. As of today, radioactive waste has not been utilized, but now they can be secondarily reprocessed, for the purpose of uranium extraction and waste utilization, since uranium prices are increasing. There is a lack of data in the literature on secondary reprocessing technologies of uranium industry wastes in Tajikistan. Each uranium tailing pile requires an individual secondary reprocessing waste technology, since they were formed as a result of different reprocessing methods (acid, soda leaching) and from different ore compositions. Their ph medium and storage conditions are different. This fact led the authors to publish the present edition of this book. The basic direction of the book is in developing manufacturing fundamentals of uranium industry waste reprocessing in Tajikistan, with specific attention on practical applications of technological investigation results. (author)

Application of a new technology for reprocessing of wastes within the framework of rehabilitation of uranium mines operated by in situ leaching - 59403

International Nuclear Information System (INIS)

Martoyan, Gagik; Nalbandyan, Garik; Gagiyan, Lavrenti; Karamyan, Gagik; Barseghyan, Artak; Brutyan, Gagik

2012-01-01

It is essential the environmentally safe industrial production of nuclear fuel especially in the case of uranium extraction by In Situ Leaching, when the environment and the deep extraction of uranium are important problems. In the presented paper it is studied the feasibility of the application of an electro-dialysis method for the deep extraction of uranium and radium from liquid (acid) streams. It is proposed to apply a new electro-hydro-metallurgical [1] extraction and refining method to ensure the necessary extraction level of elements. In the same time the new method ensures the recycling of acids used in the process. The above mentioned two different demonstrations of the new electro-hydro-metallurgical technology show that important environmental problems, such as the removal of harmful liquid-radioactive wastes, are solved in the most economical and energy efficient manner, while a new avenue has also opened for its large-scale use in mining industry. In particular, we offer this method to reprocess the huge quantity of wastes accumulated on uranium mines sites within the rehabilitation work of uranium mines operated by In Situ Leaching. A corresponding electro-hydro-metallurgical plant (mobile and stationary units) is designed for the large-scale extraction and refining of all elements from the wastes of uranium mines, which has a very high level of environmental safety, for an industry that so far has caused considerable environmental harm. The new plant design has no smokestacks, nor the emission of environmentally hazardous elements and its operation is characterized by high energy efficiency, which translates to high economy, while all materials used in the processing stages are fully reconstituted and recycled. (authors)

Uranium extraction process in a sulfuric medium by means of liquid emulsified membranes

International Nuclear Information System (INIS)

Monteillet, A.

1985-02-01

Uranium ore processing, after leaching by sulfuric acid, by liquid-liquid extraction is a rather heavy process, not suitable for small deposits. Extraction by emulsions was suggested. In this process the leachate is contacted with an oil in water type emulsion, a liquid organic membrane is formed by the continuous phase. Uranium complexes diffuse through the liquid membrane towards the dispersed aqueous phase of the emulsion (stripping solution). Uranium is recovered by breaking the emulsion. Are successively studied: development of stable emulsions, influence of emulsion composition on uranium transfer kinetics, transfer mechanisms through the membrane and modelling of kinetics data obtained in the experimental study [fr

Application of ion exchange in liquid radioactive waste management of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Pal, Puskar; Chopra, S K; Sharma, P D [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

The operation of nuclear power plants would necessarily result in generation of gaseous, liquid and solid radioactive wastes. The wastes are treated/conditioned to ensure that the permissible discharge limits laid down by Atomic Energy Regulatory Board of India are complied with. The wastes are segregated on activity levels, types of radioisotopes present and chemical nature of liquid streams. The basic philosophy of various treatment techniques is to concentrate and contain as much activity as possible. It is of utmost importance that the wastes are effectively treated by proven methods/processes. The radiochemical nature of waste generated is one of the parameters to select a treatment/conditioning method. The paper presents an outline of various processes adopted for treatment of liquid waste and ion exchange processes, their application in liquid waste management in detail. Projected quantities of liquid wastes for the current designs are included. (author). 2 tabs.

Nuclear waste viewed in a new light; a synchrotron study of uranium encapsulated in grout

Energy Technology Data Exchange (ETDEWEB)

Stitt, C.A., E-mail: Camilla.stitt@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); Hart, M., E-mail: oxford.mike@gmail.com [Diamond Light Source Limited, Harwell Science and Innovation Campus, Fermi Avenue, Didcot, Oxfordshire OX11 0QX (United Kingdom); Harker, N.J., E-mail: nicholas.harker@esrf.fr [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); Hallam, K.R., E-mail: k.r.hallam@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); MacFarlane, J., E-mail: james.macfarlane@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); Banos, A., E-mail: antonis.banos@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); Paraskevoulakos, C., E-mail: cp13846@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); Butcher, E., E-mail: ed.j.butcher@nnl.co.uk [National Nuclear Laboratory, Seascale, Cumbria CA20 1 PG (United Kingdom); Padovani, C., E-mail: cristiano.padovani@nda.gov.uk [Radioactive Waste Management Limited (formerly the Radioactive Waste Management Directorate of the UK Nuclear Decommissioning Authority), Curie Avenue, Didcot, Oxfordshire OX11 0RH (United Kingdom); Scott, T.B., E-mail: t.b.scott@bristol.ac.uk [Interface Analysis Centre, H. H. Wills Physics Laboratory, University of Bristol, Tyndall Avenue, Bristol BS8 1TL (United Kingdom)

2015-03-21

Highlights: • Unirradiated Magnox uranium was encapsulated in grout and exposed to hydrogen. • Synchrotron X-ray tomography imaged the uranium corrosion before and after exposure. • Synchrotron X-ray powder diffraction identified the corrosion products; UH{sub 3} and UO{sub 2}. • Uranium encapsulated in grout oxidised via the anoxic U + H{sub 2}O regime. • Successful in-situ, non-invasive examination of pyrophoric and radioactive material - Abstract: How do you characterise the contents of a sealed nuclear waste package without breaking it open? This question is important when the contained corrosion products are potentially reactive with air and radioactive. Synchrotron X-rays have been used to perform micro-scale in-situ observation and characterisation of uranium encapsulated in grout; a simulation for a typical intermediate level waste storage packet. X-ray tomography and X-ray powder diffraction generated both qualitative and quantitative data from a grout-encapsulated uranium sample before, and after, deliberately constrained H{sub 2} corrosion. Tomographic reconstructions provided a means of assessing the extent, rates and character of the corrosion reactions by comparing the relative densities between the materials and the volume of reaction products. The oxidation of uranium in grout was found to follow the anoxic U + H{sub 2}O oxidation regime, and the pore network within the grout was observed to influence the growth of uranium hydride sites across the metal surface. Powder diffraction analysis identified the corrosion products as UO{sub 2} and UH{sub 3}, and permitted measurement of corrosion-induced strain. Together, X-ray tomography and diffraction provide means of accurately determining the types and extent of uranium corrosion occurring, thereby offering a future tool for isolating and studying the reactions occurring in real full-scale waste package systems.

The study on process of recycling uranium in mixture of residue and liquid

International Nuclear Information System (INIS)

Zhang Jie; Shen Weiwei; Hao Jidong; Wu Jiangming

2014-01-01

The treat method of mixture of residue and liquid produced from HWR nuclear fuel chemical process using some kind of U_3O_8 powder was studied in this experiment. For recycling the uranium in mixture of residue and liquid, chemical dissolving method, washing and centrifuging method and dilute nitric acid leaching uranium method was contrasted in this test. The merit of dilute nitric acid leaching uranium method is simpler, more effective and higher uranium recycling ratio. Next, dilute nitric acid leaching uranium method was studied systematically. As a result, the main influence factors of uranium recycling ratio is dip sour degree and dip sour temperature. The influence law of factors to uranium recycling ratio and filtering effect was found out also. Along with increasing of dip sour degree and dip sour temperature, uranium recycling ratio increases and speed of filtrate increases also. At last, the process of batch treating mixture of residue and liquid was build and abundant uranium was recycled. (authors)

Passive 3D imaging of nuclear waste containers with Muon Scattering Tomography

Science.gov (United States)

Thomay, C.; Velthuis, J.; Poffley, T.; Baesso, P.; Cussans, D.; Frazão, L.

2016-03-01

The non-invasive imaging of dense objects is of particular interest in the context of nuclear waste management, where it is important to know the contents of waste containers without opening them. Using Muon Scattering Tomography (MST), it is possible to obtain a detailed 3D image of the contents of a waste container on reasonable timescales, showing both the high and low density materials inside. We show the performance of such a method on a Monte Carlo simulation of a dummy waste drum object containing objects of different shapes and materials. The simulation has been tuned with our MST prototype detector performance. In particular, we show that both a tungsten penny of 2 cm radius and 1 cm thickness, and a uranium sheet of 0.5 cm thickness can be clearly identified. We also show the performance of a novel edge finding technique, by which the edges of embedded objects can be identified more precisely than by solely using the imaging method.

Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

International Nuclear Information System (INIS)

Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

2003-01-01

The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

Improved liquid waste processing system of PWR plant

International Nuclear Information System (INIS)

Suehiro, Kazuyasu

1977-01-01

Mitsubishi Heavy Industries, Ltd. has engaged in the improvement and enhancement of waste-processing facilities for PWR power stations, and recently established the improved processing system. With this system, it becomes possible to contain radioactive waste gas semi-permanently within plants and to recycle waste liquid after the treatment, thus to make the release of radioactive wastes practically zero. The improved system has the following features, namely the recycling system is adopted, drain is separated and each separated drain is treated by specialized process, the reboiler type evaporator and the reverse osmosis equipment are used, and the leakless construction is adopted for the equipments. The radioactive liquid wastes in PWR power stations are classified into coolant drain, drain from general equipments, chemical drain and cleaning water. The outline of the improved processing system and the newly developed equipments such as the reboiler type evaporator and the reverse osmosis equipment are explained. With the evaporator, the concentration rate of waste liquid can be raised to about three times, and foaming waste can be treated efficiently. The decontamination performance is excellent. The reverse osmosis treatment is stable and reliable method, and is useful for the treatment of cleaning water. It is also effective for concentrating treatment. The unmanned automatic operation is possible. (Kako, I.)

Highly efficient extraction and selective separation of uranium (VI) from transition metals using new class of undiluted ionic liquids based on H-phosphonate anions.

Science.gov (United States)

Zarrougui, Ramzi; Mdimagh, Raouf; Raouafi, Nourreddine

2018-01-15

In this paper, we report the development of an environmental friendly process to decontaminate uranium-containing ores and nuclear wastes by using non-fluorinated ionic liquids (ILs). The main advantages of this extraction process are the absence of any organic diluent and extra extraction agents added to the organic phase. Moreover, the process is cost-effective and maybe applied as a sustainable hydrometallurgical method to recover uranium. The distribution ratio (D U ) and the extraction efficiency (%E) of uranium(VI) (UO 2 2+ ) were found to be dependent on the acidity of the aqueous phase, the extraction time, the alkyl chain length in the ILs, the concentration of the aqueous feed and molar quantity of ILs. The D U value is higher than 600 and the %E is equal to 98.6% when [HNO 3 ]=7M. The extraction reactions follows a neutral partition or ionic exchange mechanism depending on nitric acid concentration. The nature of bonding in the extracted complexes was investigated by spectroscopic techniques. The potential use of Mor 1-8 -OP for the separation of UO 2 2+ from a mixture containing transition metal ions M n+ was also examined. The UO 2 2+ ions were separated and extracted efficiently. These ILs are promising candidates for the recovery and separation of uranium. Copyright © 2017 Elsevier B.V. All rights reserved.

Characteristics and Classification of Solid Radioactive Waste From the Front-End of the Uranium Fuel Cycle.

Science.gov (United States)

Liu, Xinhua; Wei, Fangxin; Xu, Chunyan; Liao, Yunxuan; Jiang, Jing

2015-09-01

The proper classification of radioactive waste is the basis upon which to define its disposal method. In view of differences between waste containing artificial radionuclides and waste with naturally occurring radionuclides, the scientific definition of the properties of waste arising from the front end of the uranium fuel cycle (UF Waste) is the key to dispose of such waste. This paper is intended to introduce briefly the policy and practice to dispose of such waste in China and some foreign countries, explore how to solve the dilemma facing such waste, analyze in detail the compositions and properties of such waste, and finally put forward a new concept of classifying such waste as waste with naturally occurring radionuclides.

Recent studies of uranium and plutonium chemistry in alkaline radioactive waste solutions

International Nuclear Information System (INIS)

King, William D.; Wilmarth, William R.; Hobbs, David T.; Edwards, Thomas B.

2008-01-01

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

Radioactive liquid waste processing system

International Nuclear Information System (INIS)

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

1996-01-01

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

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.)

THE RADIOECOLOGICAL CHARACTERISTIC OF THE RADIOACTIVE WASTE DISPOSAL SITES OF URANIUM PRODUCTION AND RECOMMENDATIONS FOR THEIR IMPROVEMENT

Directory of Open Access Journals (Sweden)

O. L. Ten

2015-01-01

Full Text Available This article deals with the issues of management of the uranium industry wastes at the territory of the Republic of Uzbekistan. The characteristics of the main uranium deposits and mining methods are given. The environmental problems associated with presence of large areas occupied by the uranium industry wastes and with the influence of natural processes on the integrity of such storage places are discussed. The author gives a generalized description of the measures on improving radiation situation in the ecosystems at contaminated territories and facilities of the country. The efforts of the State in relation to the remediation of contaminated land on dangerous facilities are described. The main directions of responsibility and control of radiation safety at the territories containing the uranium industry wastes are indicated. Based on the analysis, the author highlights promising areas of the state policy to improve the system of protective actions to reduce existing or unregulated radiation risks. The main tasks in the field of protection of public health and the environment are defined in the paper.

Method of processing radioactive liquid wastes by solidification with cement

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki.

1975-01-01

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

Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification

NARCIS (Netherlands)

E. de Blois (Erik); H.S. Chan (Ho Sze); K. Roy (Kamalika); E.P. Krenning (Eric); W.A.P. Breeman (Wouter)

2011-01-01

textabstractPET with68Ga from the TiO2- or SnO2- based68Ge/68Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity (68Ge vs.68Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts

Measures for waste water management from recovery processing of Zhushanxia uranium deposit

International Nuclear Information System (INIS)

Liu Yaochi; Xu Lechang

2000-01-01

Measures for waste water management from recovery processing of Zhushanxia uranium deposit of Wengyuan Mine is analyzed, which include improving process flow, recycling process water used in uranium mill as much as possible and choosing a suitable disposing system. All these can decrease the amount of waste water, and also reduce costs of disposing waste water and harm to environment

Uranium-scintillator device

International Nuclear Information System (INIS)

Smith, S.D.

1979-01-01

The calorimeter subgroup of the 1977 ISABELLE Summer Workshop strongly recommended investigation of the uranium-scintillator device because of its several attractive features: (1) increased resolution for hadronic energy, (2) fast time response, (3) high density (i.e., 16 cm of calorimeter per interaction length), and, in comparison with uranium--liquid argon detectors, (4) ease of construction, (5) simple electronics, and (6) lower cost. The AFM group at the CERN ISR became interested in such a calorimeter for substantially the same reasons, and in the fall of 1977 carried out tests on a uranium-scintillator (U-Sc) calorimeter with the same uranium plates used in their 1974 studies of the uranium--liquid argon (U-LA) calorimeter. The chief disadvantage of the scintillator test was that the uranium plates were too small to fully contain the hadronic showers. However, since the scintillator and liquid argon tests were made with the plates, direct comparison of the two types of devices could be made

System for the hydrogeologic analysis of uranium mill waste disposal sites

International Nuclear Information System (INIS)

Osiensky, J.L.

1983-01-01

Most of the uranium mill wastes generated before 1977 are stored in unlined tailings ponds. Seepage from some of these ponds has been of sufficient severity that the US Nuclear Regulatory Commission (NRC) has required the installation of withdrawal wells to remove the contaminated groundwater. Uranium mill waste disposal facilities typically are located in complex hydrogeologic environments. This research was initiated in 1980 to analyze hydrogeologic data collected at seven disposal sites in the US that have experienced problems with groundwater contamination. The characteristics of seepage migration are site specific and are controlled by the hydrogeologic environment in the vicinity of each tailings pond. Careful monitoring of most seepage plumes was not initiated until approximately 1977. These efforts were accelerated as a consequence of the uranium Mill Tailings Act of 1979. Some of the data collected at uranium mill waste disposal sites in the past are incomplete and some were collected by methods that are outdated. Data frequently were collected in sequences which disrupted the continuity of the hydrogeologic analysis and decreased the effectiveness of the data collection programs. Evaluation of data collection programs for seven uranium mill waste disposal sites in the US has led to the development and presentation herein of a system for the hydrogeologic analysis of disposal sites

Healthcare liquid waste management.

Science.gov (United States)

Sharma, D R; Pradhan, B; Pathak, R P; Shrestha, S C

2010-04-01

The management of healthcare liquid waste is an overlooked problem in Nepal with stern repercussions in terms of damaging the environment and affecting the health of people. This study was carried out to explore the healthcare liquid waste management practices in Kathmandu based central hospitals of Nepal. A descriptive prospective study was conducted in 10 central hospitals of Kathmandu during the period of May to December 2008. Primary data were collected through interview, observation and microbiology laboratory works and secondary data were collected by records review. For microbiological laboratory works,waste water specimens cultured for the enumeration of total viable counts using standard protocols. Evidence of waste management guidelines and committees for the management of healthcare liquid wastes could not be found in any of the studied hospitals. Similarly, total viable counts heavily exceeded the standard heterotrophic plate count (p=0.000) with no significant difference in such counts in hospitals with and without treatment plants (p=0.232). Healthcare liquid waste management practice was not found to be satisfactory. Installation of effluent treatment plants and the development of standards for environmental indicators with effective monitoring, evaluation and strict control via relevant legal frameworks were realized.

Removal of actinide elements from liquid scintillation cocktail wastes using liquid-liquid extraction and demulsification techniques

International Nuclear Information System (INIS)

Foltz, K.; Landsberger, S.; Srinivasan, B.; Vandegrift, G.F.

1994-01-01

For many years liquid scintillation cocktail (LSC) wastes have been generated and stored at Argonne National Laboratory (ANL). These wastes are stored in thousands of 10--20 m scintillation vials, many of which contain elements with Z > 88. Because storage space is limited, disposal of this waste is pressing. These wastes could be commercially incinerated if the radionuclides with Z>88 are reduced to sufficiently low levels. However, there is currently no deminimus level for these radionuclides, and separation techniques are still being tested. The University of Illinois is conducting experiments to separate radionuclides with Z > 88 from simulated LSC wastes by using liquid-liquid extraction (LLX) and demulsification techniques. The actinide elements are removed from the LSC by extraction into an aqueous phase after the cocktail has been demulsified. The aqueous and organic phases are separated and the organic phase, now free from radionuclides with Z > 88, can be sent to a commercial incineration facility. The aqueous phase may be treated and disposed of using existing techniques. The LLX separation techniques used solutions of sodium oxalate, aluminum nitrate, and tetrasodium EDTA at varying concentrations. These extractants were mixed with the simulated waste in a 1:1 volume ratio. Using 1.0M Na 4 EDTA salt solutions, decontamination ratios as high as 230 were achieved

Immobilization of organic liquid wastes

International Nuclear Information System (INIS)

Greenhalgh, W.O.

1985-01-01

This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3-month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

Study of immobilization of waste from treatment of acid waters of a uranium mining facility

International Nuclear Information System (INIS)

Goda, R.T.; Oliveira, A.P. de; Silva, N.C. da; Villegas, R.A.S.; Ferreira, A.M.

2017-01-01

This study aimed to produce scientific and technical knowledge aiming at the development of techniques to immobilize the waste generated in the treatment of acid waters in the UTM-INB Caldas uranium mining and processing facility using Portland cement. This residue (calcium diuranate - DUCA) contains uranium compounds and metal hydroxides in a matrix of calcium sulfate. It is observed that this material, in contact with the lake of acid waters of the mine's own pit, undergoes resolubilization and, therefore, changes the quality of the acidic water contained therein, changing the treatment parameters. For the study of immobilization of this residue, the mass of water contained in both the residue deposited in the pit of the mine and in the pulp resulting from the treatment of the acid waters was determined. In addition, different DUCA / CEMENT / WATER ratios were used for immobilization and subsequent mechanical strength and leaching tests. The results showed that in the immobilized samples with 50% cement mass condition, no uranium was detected in the leaching tests, and the mechanical strength at compression was 9.4 MPa, which indicates that more studies are needed, but indicate a good capacity to immobilize uranium in cement

Liquid-liquid extraction of uranium (VI) using Cyanex 272 in kerosene from sodium salicylate medium

International Nuclear Information System (INIS)

Kamble, Pravin N.; Mohite, Baburao S.; Suryavanshi, Vishal J.; Salunkhe, Suresh T.

2015-01-01

Liquid-liquid extraction of uranium (VI) from sodium salicylate media using Cyanex 272 in kerosene has been carried out. Uranium (VI) was quantitatively extracted from 1x10 -4 M sodium salicylate with 5x10 -4 M Cyanex 272 in kerosene. It was stripped quantitatively from the organic phase with 4M HCl and determined spectrophotometrically with arsenazo(III) at 600 nm. The effects of concentrations of sodium salicylate, metal ions and strippants have been studied. Separation of uranium (VI) from other elements was achieved from binary as well as from multicomponent mixtures. The method is simple, rapid and selective with good reproducibility (approximately ±2%). (author)

Liquid radioactive wastes from hospitals by polymeric membrane

International Nuclear Information System (INIS)

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

1998-01-01

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

Experimental tests performed with liquid waste contained in the tank F-710/D at EUREX plant

International Nuclear Information System (INIS)

Gasso, G.; Momo, S.; Pietrelli, L.; Troiani, F.

1989-11-01

In this report the result of experimental test performed with real liquid waste earning from reprocessing of MTR nuclear fuel is reported. The aim of the research is to separate the actinides and long-lived radioactive fission products from bulk salt matrix of HLW. Taking into account the chemical and radiochemical composition of the liquid waste, process based on the chemical precipitation and/or adsorption were studied by using the radioactive waste sampled from the tank. The results show that decontamination factors of 100, 1000, 5000 were obtained for Sr, Cs and Pu respectively. (author)

Development of a test system for high level liquid waste partitioning

Directory of Open Access Journals (Sweden)

Duan Wu H.

2015-01-01

Full Text Available The partitioning and transmutation strategy has increasingly attracted interest for the safe treatment and disposal of high level liquid waste, in which the partitioning of high level liquid waste is one of the critical technical issues. An improved total partitioning process, including a tri-alkylphosphine oxide process for the removal of actinides, a crown ether strontium extraction process for the removal of strontium, and a calixcrown ether cesium extraction process for the removal of cesium, has been developed to treat Chinese high level liquid waste. A test system containing 72-stage 10-mm-diam annular centrifugal contactors, a remote sampling system, a rotor speed acquisition-monitoring system, a feeding system, and a video camera-surveillance system was successfully developed to carry out the hot test for verifying the improved total partitioning process. The test system has been successfully used in a 160 hour hot test using genuine high level liquid waste. During the hot test, the test system was stable, which demonstrated it was reliable for the hot test of the high level liquid waste partitioning.

Liquidation of wastes as tuition topic

International Nuclear Information System (INIS)

Kolar, K.; Hysplerova, L.; Holy, I.

1999-01-01

Authors deal in this paper with tuition project aimed on the liquidation of wastes. Structure of project includes next thematic complex: classification of inorganic and organic wastes; characterization of wastes and proposition for their liquidation (detoxication) or recyclation; chemical (physical) nature of neutralize of inorganic and organic wastes; general method of neutralize of wastes; analytical methods necessary for control of course of neutralize (detoxication) of wastes. This tuition project allows for students to know manipulation with wastes and methods of their liquidation from ecologic point of view

Development of Characterization Protocol for Mixed Liquid Radioactive Waste Classification

International Nuclear Information System (INIS)

Norasalwa Zakaria; Syed Asraf Wafa; Wo, Y.M.; Sarimah Mahat; Mohamad Annuar Assadat Husain

2017-01-01

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

Development of characterization protocol for mixed liquid radioactive waste classification

Energy Technology Data Exchange (ETDEWEB)

Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my [Waste Technology Development Centre, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wafa, Syed Asraf [Radioisotop Technology and Innovation, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wo, Yii Mei [Radiochemistry and Environment, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Mahat, Sarimah [Material Technology Group, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

2015-04-29

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

Determination of uranium in liquid samples

International Nuclear Information System (INIS)

Macefat, Martina R.; Grahek, Zeljko; Ivsic, Astrid G.

2008-01-01

Full text: Uranium is a natural occurring radionuclide and the first member of natural radioactive chains which makes its determination in natural materials interesting from geochemical and radioecological aspect. It can be quantitatively determined as element and/or its radioisotopes by different spectrometric methods (ICP-MS, spectrophotometry, alpha spectrometry). It is necessary to develop inexpensive, rapid and sensitive methods for the routine analysis. Therefore, in this paper, development of a new method for the isolation of uranium from liquid samples and subsequent determination by spectrophotometry and ICP-MS will be described. It is possible to isolate uranium from drinking and seawater using extraction chromatography or mixed solvent ion exchange. Uranium can be strongly bound on the TRU extraction chromatographic resin from nitric acid (chemical recovery is 100%) and can be separated from other interfering elements, while separation from thorium, which can be also strongly bound on this resin, is possible with hydrochloric acid. It is also possible to separate uranium from thorium on the anion exchanger Amberlite CG-400 (NO 3 - form) because uranium is much more weakly bound on this exchanger from alcoholic solutions of nitric acid. After the separation uranium is determined by ICP-MS and by spectrophotometric method with arsenazo III (λ max =652 nm). Developed method enables selection of the optimal mode of isolation for the given purposes. (author)

Method for ultimate disposition of borate containing radioactive wastes by vitrification

International Nuclear Information System (INIS)

Bege, D.; Faust, H.J.; Puthawala, A.; Stunkel, H.

1984-01-01

Method for the ultimate disposition of radioactive wastes by vitrification, in which weak to medium radioactive waste concentrates from borate-containing radioactive liquids are mixed with added glass-forming materials, maximally in a ratio of 1:3, and the mixture heated to obtain a glass-forming melt

Study on direct dissolution of U-10Zr alloy and distribution of uranium and zirconium in liquid cadmium

International Nuclear Information System (INIS)

Ye Yuxing; Gao Yuan

1997-09-01

The effect of dissolution time, temperature, total surface area of U-10Zr alloy pellets and stirring on the dissolution and dissolution rate of uranium in liquid cadmium were studied. Cadmium containing U and Zr dissolved from U-10Zr alloy at 475 degree C and 500 degree C respectively was analyzed with electron microanalyzer. The experimental results show that at 400 degree and 500 degree C with the stirring rate of some 150 r/min, the solubilities of uranium in liquid cadmium are 0.4% and 2.2%, respectively. At the first 30 min, the dissolution rates of U-10Zr alloy pellets are 0.05 g/(cm 2 ·h) and 0.32 g/(cm 2 ·h), respectively. The suitable dissolution conditions for U-10Zr alloy pellets in liquid cadmium (the ratio of the mass of liquid cadmium to that of the pellets ≅7) are: temperature, about 480 degree C; stirring rate, about 150 r/min; dissolution time, 4 h. The distribution of uranium and zirconium in cadmium is homogeneous

Characterization of radioactive organic liquid wastes

International Nuclear Information System (INIS)

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

2014-10-01

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

Non-polluting treatment of uranium effluents from the alkaline digestion of an uranium ore containing sulfur

International Nuclear Information System (INIS)

Berger, Bernard.

1978-01-01

New non-polluting process for treating uranium effluents from the alkaline digestion of an uranium ore containing sulphur, which makes it possible (a) to extract and obtain relatively pure uranium and (b) to process the digestion liquor freed from the uranium and containing in an aqueous solution a mixture of alkaline carbonate and/or bicarbonate and sodium sulphate, consisting in the selective extraction of the sodium sulphate present and the recycling of the liquor free of SO 4 = ions, containing in solution the sole carbonates and/or bicarbonates involved, towards the digestion of the ore [fr

Uranium recovery from waste of the nuclear fuel cycle plants at IPEN-CNEN/SP, Brazil

Energy Technology Data Exchange (ETDEWEB)

Freitas, Antonio A.; Ferreira, Joao C.; Zini, Josiane; Scapin, Marcos A.; Carvalho, Fatima Maria Sequeira de, E-mail: afreitas@ipen.b, E-mail: jcferrei@ipen.b, E-mail: jzini@ipen.b, E-mail: mascapin@ipen.b, E-mail: fatimamc@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Sodium diuranate (DUS) is a uranium concentrate produced in monazite industry with 80% typical average grade of U{sup 3}O{sup 8}, containing sodium, silicon, phosphorus, thorium and rare earths as main impurities. Purification of such concentrate was achieved at the nuclear fuel cycle pilot plants of uranium at IPEN by nitric dissolution and uranium extraction into an organic phase using TBP/Varsol, while the aqueous phase retains impurities and a small quantity of non extracted uranium; both can be recovered later by precipitation with sodium hydroxide. Then the residual sodium diuranate goes to a long term storage at a safeguards deposit currently reaching 20 tonnes. This work shows how uranium separation and purification from such bulk waste can be achieved by ion exchange chromatography, aiming at decreased volume and cost of storage, minimization of environmental impacts and reduction of occupational doses. Additionally, the resulting purified uranium can be reused in nuclear fuel cycle.(author)

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.)

The Effectivity of Marine Bio-activator and Surimi Liquid Waste Addition of Characteristics Liquid Organic Fertilizer from Sargassum sp.

Directory of Open Access Journals (Sweden)

Putri Wening Ratrinia

2017-02-01

Full Text Available AbstractOrganic fertilizer is highly recommended for soil and plant because it can improve the productivity and repair physical, chemical, and biological of soil. Sargassum sp. and surimi liquid wastes contain organic matter and nutrient needed by plants and soils. The addition of marine bio-activator which contains bacterial isolates from litter mangrove serves to accelerate the composting time and increases the activity of microorganisms in the decomposition process. The purpose of this study was to determine optimum time and the best formulation of decomposition process organic fertilizer. Raw materials used a waste of seaweed Sargassum sp., marine bio-activator and surimi liquid waste from catfish (Clarias sp.. The research was conducted six treatments control, Sargassum sp. + marine bio-activator, surimi liquid waste , Sargassum sp. + marine bio-activator + surimi liquid waste 80%, 90%, 100%. All treatments were fermented for 9 days and analysed the C-organic, total N, C/N ratio, P2O5, K2O on days 0, 3, 6 and 9. The results showed the optimum fermentation period was on the 6th day. The most optimum concentration of surimi liquid waste added was at a concentration of 90%, with characteristics of the products was C-organic 0.803±0.0115%, total N 740.063±0.0862 ppm, C/N ratio 10.855±0.1562, P2O5 425.603±0.2329 ppm, K2O 2738.627±0.2836 ppm.

The Effectivity of Marine Bio-activator and Surimi Liquid Waste Addition of Characteristics Liquid Organic Fertilizer from Sargassum sp.

Directory of Open Access Journals (Sweden)

Putri Wening Ratrinia

2016-12-01

Full Text Available Organic fertilizer is highly recommended for soil and plant because it can improve the productivity and repair physical, chemical, and biological of soil. Sargassum sp. and surimi liquid wastes contain organic matter and nutrient needed by plants and soils. The addition of marine bio-activator which contains bacterial isolates from litter mangrove serves to accelerate the composting time and increases the activity of microorganisms in the decomposition process. The purpose of this study was to determine optimum time and the best formulation of decomposition process organic fertilizer. Raw materials used a waste of seaweed Sargassum sp., marine bio-activator and surimi liquid waste from catfish (Clarias sp.. The research was conducted six treatments control, Sargassum sp. + marine bio-activator, surimi liquid waste , Sargassum sp. + marine bio-activator + surimi liquid waste 80%, 90%, 100%. All treatments were fermented for 9 days and analysed the C-organic, total N, C/N ratio, P2 O5 , K2 O on days 0, 3, 6 and 9. The results showed the optimum fermentation period was on the 6th day. The most optimum concentration of surimi liquid waste added was at a concentration of 90%, with characteristics of the products was C-organic 0.803 ± 0.0115 %, total N 740.063 ± 0.0862 ppm, C/N ratio 10.855 ± 0.1562, P2 O5 425.603 ± 0.2329 ppm, K2 O 2738.627 ± 0.2836 ppm.

Separation of transuranium elements and fission products from medium activity aqueous liquid wastes

International Nuclear Information System (INIS)

Gompper, K.; Kunze, S.; Eden, G.; Loesch, G.; Zemski, C.

1986-01-01

In the course of work performed between January 1981 and June 1985 on the separation of TRU elements and fission products three liquid alpha containing waste streams were treated: - medium level waste solutions, - waste solutions from the acid digestion of burnable alpha containing solid residues, - waste solutions from mixed oxide fuel element fabrication. The method of separation was initially developed and optimized with simulating substances. Subesequently it was tested with real waste solutions

Some potential strategies for the treatment of waste uranium metal and uranium alloys

International Nuclear Information System (INIS)

Burns, C.J.; Frankcom, T.M.; Gordon, P.L.; Sauer, N.N.

1993-01-01

Large quantities of uranium metal chips and turnings stored throughout the DOE Complex represent a potential hazard, due to the reactivity of this material toward air and water. Methods are being sought to mitigate this by conversion of the metal, via room temperature solutions routes, to a more inert oxide form. In addition, the recycling of uranium and concomitant recovery of alloying metals is a desirable goal. The emphasis of the authors' research is to explore a variety of oxidation and reduction pathways for uranium and its compounds, and to investigate how these reactions might be applied to the treatment of bulk wastes

Uranium mining, atomic weapons testing, nuclear waste storage: A global survey. World Uranium Hearing grey book 1992

International Nuclear Information System (INIS)

Krumbholz, E.; Kressing, F.

1992-09-01

The first edition of the 'World Uranium Hearing Grey Book' for the World Uranium Hearing in Salzburg, 13-19 September 1992 is meant to be a reference for people involved in the World Uranium Hearing. It is mostly made up to country by country surveys giving background information on the testimonies presented at the Hearing, and on many more cases. Included are two short articles: One on 'nukespeak' to make the reader aware of how the language of the nuclear industry influences our speaking and thinking; and an article on the wastes produced by uranium mines. Due to limited time and resources this documentation is not complete. Many questions remain. For example, information is rare about conditions in Eastern Europe. Also, some countries are given much more space than others, which does not indicate importance or seriousness of implications of uranium mining, weapons testing or nuclear waste storage in this particular country. (orig./HP)

Environmental performance evaluation of waste management system of Uranium Concentrated Unit in Caetite city, Bahia State - Brazil

International Nuclear Information System (INIS)

Araujo, Valeska P.; Fernandes, Horst M.; Gomiero, Luis Alberto

2005-01-01

The mining/milling activities have the potential to cause risks to the human health and to the environment. In uranium mining, besides inherent damages due to any mining activities there are radiological risks, that may be incurred even in short as in long terms. The large volumes of low activity mining/milling residues produced, are the great challenge in the waste management. Nowadays, the whole Brazilian uranium production come from Uranium Concentrated Unit (URA), a facility operated by Brazilian Nuclear Industry and located at a semi-arid region, in the Caetite city, Bahia state. This Unit is composed by a open pit mine and a milling facility. The present work assess the URA waste management system, the procedures adopted, focused on its environmental performance. It was observed that the waste management system is efficient in the control of the environmental impacts, however improvement chances are detected and a better performance may be reached. Concerning the liquids wastes, it was observed that the storage systems were not projected adequately. The storage capacity was not enough to support a intense rainfall period causing a overflow to the environment. In URA activities there is no radiological risk to the public, but its necessary to improve long term actions, constraints for the post-closure phase, e.g., appropriate institutional controls, restrictions on land use. Finally, it is advisable to introduce a Environmental Management System (EMS) for the whole facility. (author)

The uranium production cycle and the environment. Proceedings

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-04-01

issues and performance in several countries; problems of acid rock drainage at uranium mine sites; national examples of environmental impact assessment in the uranium industry; natural background radioactivity of the earth's surface at and away from uranium production facilities; technology and the uranium industry; environmental design aspects of uranium mines and mills; design, management and closure of uranium tailings facilities; national experience with management and disposal of liquid and solid wastes from uranium mining and milling; national experience with treatment and restoration of waste water from both conventional and in situ leach uranium mining; safe management of radioactive waste from mining and milling activities; evolving regulation and rehabilitation standards for uranium recovery operations; national experience with regulation of uranium production. This publication contains 71 individual papers presented at the symposium. Each of the papers was indexed separately.

The uranium production cycle and the environment. Proceedings

International Nuclear Information System (INIS)

2002-01-01

and performance in several countries; problems of acid rock drainage at uranium mine sites; national examples of environmental impact assessment in the uranium industry; natural background radioactivity of the earth's surface at and away from uranium production facilities; technology and the uranium industry; environmental design aspects of uranium mines and mills; design, management and closure of uranium tailings facilities; national experience with management and disposal of liquid and solid wastes from uranium mining and milling; national experience with treatment and restoration of waste water from both conventional and in situ leach uranium mining; safe management of radioactive waste from mining and milling activities; evolving regulation and rehabilitation standards for uranium recovery operations; national experience with regulation of uranium production. This publication contains 71 individual papers presented at the symposium. Each of the papers was indexed separately

Pre-concentration of uranium from water samples by dispersive liquid-liquid micro-extraction

Energy Technology Data Exchange (ETDEWEB)

Khajeh, Mostafa; Nemch, Tabandeh Karimi [Zabol Univ. (Iran, Islamic Republic of). Dept. of Chemistry

2014-07-01

In this study, a simple and rapid dispersive liquid-liquid microextraction (DLLME) was developed for the determination of uranium in water samples prior to high performance liquid chromatography with diode array detection. 1-(2-pyridylazo)-2-naphthol (PAN) was used as complexing agent. The effect of various parameters on the extraction step including type and volume of extraction and dispersive solvents, pH of solution, concentration of PAN, extraction time, sample volume and ionic strength were studied and optimized. Under the optimum conditions, the limit of detection (LOD) and preconcentration factor were 0.3 μg L{sup -1} and 194, respectively. Furthermore, the relative standard deviation of the ten replicate was <2.6%. The developed procedure was then applied to the extraction and determination of uranium in the water samples.

Pre-concentration of uranium from water samples by dispersive liquid-liquid micro-extraction

International Nuclear Information System (INIS)

Khajeh, Mostafa; Nemch, Tabandeh Karimi

2014-01-01

In this study, a simple and rapid dispersive liquid-liquid microextraction (DLLME) was developed for the determination of uranium in water samples prior to high performance liquid chromatography with diode array detection. 1-(2-pyridylazo)-2-naphthol (PAN) was used as complexing agent. The effect of various parameters on the extraction step including type and volume of extraction and dispersive solvents, pH of solution, concentration of PAN, extraction time, sample volume and ionic strength were studied and optimized. Under the optimum conditions, the limit of detection (LOD) and preconcentration factor were 0.3 μg L -1 and 194, respectively. Furthermore, the relative standard deviation of the ten replicate was <2.6%. The developed procedure was then applied to the extraction and determination of uranium in the water samples.

Waste analysis plan for 222-S dangerous and mixed waste storage area

International Nuclear Information System (INIS)

Warwick, G.J.

1994-01-01

The 222-S Laboratory Complex, in the southeast corner of the 200 West Area, consists of the 222-S Laboratory, the 222-SA Standards Laboratory, and several ancillary facilities. Currently, 222-S Laboratory activities are in supporting efforts to characterize the waste stored in the 200 Areas single shell and double shell tanks. Besides this work, the laboratory also provides analytical services for waste-management processing plants, Tank Farms, B Plant, 242-A Evaporator Facility, Plutonium-Uranium Extraction Plant, Plutonium Finishing Plant, Uranium-Oxide Plant, Waste Encapsulation Storage Facility, environmental monitoring and surveillance programs, and activities involving essential materials and research and development. One part of the 222-SA Laboratory prepares nonradioactive standards for the 200 Area laboratories. The other section of the laboratory is used for cold (nonradioactive) process development work and standards preparation. The 219-S Waste Handling Facility has three storage tanks in which liquid acid waste from 222-S can be received, stored temporarily, and neutralized. From this facility, neutralized waste, containing radionuclides, is transferred to the Tank Farms. A 700-gallon sodium-hydroxide supply tank is also located in this facility. This plan provides the methods used to meet the acceptance criteria required by the 204-AR Waste Receiving Facility

Transportation of liquid mixed waste in the US: Is it really a problem?

International Nuclear Information System (INIS)

Chakraborti, S.; DeBiase, T.

1993-01-01

The transportation of liquid radioactive wastes has often been perceived to be a problem because of the potential consequences from hypothetical accident scenarios and the difficulties that may be encountered in the handling and containment of liquids. This paper focuses specifically to determine if the transportation of these wastes are severely restricted by the regulations. The paper also compares current practices for the transportation of liquid mixed waste in the US with that of France to provide an international perspective on the issue. The review of the regulations and current practices shows that the transportation of liquid mixed waste is by no means prohibited, and also that the majority of the regulations do not impose any additional restrictions because of the physical form of the waste. Rather, the selection of an authorized package primarily depends on the quantity of radioactivity and the specific radionuclides involved. Although the selection process for an authorized package for liquid mixed wastes is fairly straightforward, it seems that the difficulties in transporting liquid mixed waste can be attributed to the lack of readily available Type A packages designed for transporting liquids

VUJE experience with cementation of liquid and wet radioactive waste

International Nuclear Information System (INIS)

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

2011-01-01

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

Adsorption equilibrium of uranium on iron oxyhydroxide-PVA hydrogel spheres

Energy Technology Data Exchange (ETDEWEB)

Santos, Armindo; Campos, Victor B.; Ribeiro, Luciana S.; Escanio, Camila A.; Silva, Edilaine F.; Oliveira, Felipe W., E-mail: santosa@cdtn.br, E-mail: vbc@cdtn.br, E-mail: lsr@cdtn.br, E-mail: cae@cdtn.br, E-mail: efd@cdtn.br, E-mail: fwfo@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Uranium and its compounds are considered strategic mineral resources due to its usage as an energy source and war material. They are harmful to human health. Thus, liquid waste containing low uranium content (≤100 mgU/L), from the mining and/or uranium reprocessing plants or even of the research center activities require the development of methods for their treatment, in a way to reduce its content to 15 μgU/L. Adsorption is one of these methods; it requires the synthesis of preferably spherical adsorbents, chemically and physically stable and with high adsorptive capacity. The sol-gel process can synthesize adsorbents having such characteristics, prioritizing the nanostructuring of iron oxyhydroxide in a hydrophilic PVA (polyvinyl alcohol) polymer network, which had an accessible pore structure (micro-, meso- and macropores + macroholes). We successfully obtained iron-PVA hydrogel spheres with (3433 ± 63 μm) and without (2833 ± 69 μm) macroholes. Both types of spheres have good mechanical strength and chemical stability in the 2-9 pH range. Adsorptive capacity: 413.22 mgU/g (with macroholes; Freundlich model) and 249.38 mgU/g (without macroholes; Langmuir and Freundlich models), at pH 5-6, 30 °C, and 6 h. With 280 mL of with-macrohole hydrogel spheres, we can treat 1 L of liquid waste (100 mgU/L) and reduce uranium content to 20 μgU/L. (author)

Adsorption equilibrium of uranium on iron oxyhydroxide-PVA hydrogel spheres