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Sample records for liquid waste solution

  1. China's Scientific Investigation for Liquid Waste Treatment Solutions

    International Nuclear Information System (INIS)

    Liangjin, B.; Meiqiong, L.; Kelley, D.

    2006-01-01

    Post World War II created the nuclear age with several countries developing nuclear technology for power, defense, space and medical applications. China began its nuclear research and development programs in 1950 with the establishment of the China Institute of Atomic Energy (CIAE) located near Beijing. CIAE has been China's leader in nuclear science and technical development with its efforts to create advanced reactor technology and upgrade reprocessing technology. In addition, with China's new emphasis on environmental safety, CIAE is focusing on waste treatment options and new technologies that may provide solutions to legacy waste and newly generated waste from the full nuclear cycle. Radioactive liquid waste can pose significant challenges for clean up with various treatment options including encapsulation (cement), vitrification, solidification and incineration. Most, if not all, nuclear nations have found the treatment of liquids to be difficult, due in large part to the high economic costs associated with treatment and disposal and the failure of some methods to safely contain or eliminate the liquid. With new environmental regulations in place, Chinese nuclear institutes and waste generators are beginning to seek new technologies that can be used to treat the more complex liquid waste streams in a form that is safe for transport and for long-term storage or final disposal. [1] In 2004, CIAE and Pacific Nuclear Solutions, a division of Pacific World Trade, USA, began discussions about absorbent technology and applications for its use. Preliminary tests were conducted at CIAE's Department of Radiochemistry using generic solutions, such as lubricating oil, with absorbent polymers for solidification. Based on further discussions between both parties, it was decided to proceed with a more formal test program in April, 2005, and additional tests in October, 2005. The overall objective of the test program was to apply absorbent polymers to various waste streams

  2. Radiolytic gas formation in high-level liquid waste solutions

    International Nuclear Information System (INIS)

    Brodda, B.-G.; Dix, Siegfried; Merz, E.R.

    1989-01-01

    High-level fission product waste solutions originating from the first-cycle raffinate stream of spent fast breeder reactor fuel reprocessing have been investigated gas chromatographically for their radiolytic and chemical gas production. The solutions showed considerable formation of hydrogen, carbon dioxide and dinitrogen oxide, whereas atmospheric oxygen was consumed completely within a short time. In particular, carbon dioxide resulted from the radiolytic degradation of entrained organic solvent. After nearly complete degradation of the organic solvent, the influence of hydrazine and nitrogen dioxide on hydrogen formation was investigated. Hydrazinium hydroxide led to the formation of dinitrogen oxide and nitrogen. After 60 d, the concentration of dinitrogen oxide had reduced to zero, whereas the amount of nitrogen formed had reached a maximum. This may be explained by simultaneous chemical and radiolytic reactions leading to the formation of dinitrogen oxide and nitrogen and photolytic fission of dinitrogen oxide. Addition of sodium nitrite resulted in the rapid formation of dinitrogen oxide. The rate of hydrogen production was not changed significantly after the addition of hydrazine or nitrite. The results indicate that under normal operating conditions no dangerous hydrogen radiolysis yields should develop in the course of reprocessing and high-level liquid waste tank storage. Organic entrainment may lead to enhanced radiolytic decomposition and thus to considerable hydrogen production rates and pressure build-up in closed systems. (author)

  3. A novel Canadian solution for processing and disposal of mixed liquid wastes

    International Nuclear Information System (INIS)

    Suryanarayan, S.; Husain, A.; Husain, S.; Grey, M.; Elwood, C.; White, T.; Wigle, K.

    2011-01-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  4. A novel Canadian solution for processing and disposal of mixed liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Suryanarayan, S.; Husain, A. [Kinectrics Inc., Toronto, ON (Canada); Husain, S.; Grey, M. [Candesco, Toronto, ON (Canada); Elwood, C.; White, T.; Wigle, K. [Bruce Power, Tiverton, ON (Canada)

    2011-07-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  5. Pore solution chemistry of simulated low-level liquid waste incorporated in cement grouts

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1995-12-01

    Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50 degree C and 90 degree C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na + , P0 4 3- , N0 2 - , NO 3 - and OH - . The other is carbonated and contains the species, Na + , PO 4 3- , NO 2 - , NO 3 - , and CO 3 2- . In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH - and Na + ion concentrations. In contrast high concentration of N0 2 - and N0 3 - were retained in the pore solution indefinitely

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-12

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

  7. 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.)

  8. Corrosivity of solutions from evaporation of radioactive liquid wastes. Final report

    International Nuclear Information System (INIS)

    Payer, H.; Kolic, E.S.; Boyd, W.K.

    1977-01-01

    New double-shell storage tanks are constructed with ASTM A-516 Grade 65 steel. This study had two main objectives: To characterize the corrosivity of synthetic nonradioactive terminal waste solutions to ASTM A-516 Grade 65 steel and to determine the severity of stress-corrosion cracking of carbon steel in terminal waste solutions. The information developed provides guidance in the characterization of the aggressiveness of actual terminal liquors and in the design and operation of fail-safe tanks. Corrosion behavior was measured over a range of oxidizing conditions by the potentiodynamic polarization technique. Oxidizing conditions in a solution likely to promote general corrosion, pitting or stress-corrosion cracking (SCC) were identified. Absolute stress-corrosion cracking susceptibility was determined by constant strain rate procedure for ASTM A-516 Grade 65 steel for conditions identified by polarization experiments as likely to promote SCC. Based on the results of this study, terminal waste storage tanks are safe from stress-corrosion cracking under freely corroding conditions. Corrosion potential of steel in solutions within anticipated compositions is at the positive end of the critical range for stress-corrosion cracking, and no conditions were observed which would lower the potential to more negative values within the cracking range under freely corroding conditions. Measurement of corrosion potential and hydroxide concentration provides a means to extend these results to compositions outside of the composition range studied

  9. 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.)

  10. 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.)

  11. Liquid scintillation solutions

    International Nuclear Information System (INIS)

    Long, E.C.

    1976-01-01

    The liquid scintillation solution described includes a mixture of: a liquid scintillation solvent, a primary scintillation solute, a secondary scintillation solute, a variety of appreciably different surfactants, and a dissolving and transparency agent. The dissolving and transparency agent is tetrahydrofuran, a cyclic ether. The scintillation solvent is toluene. The primary scintillation solute is PPO, and the secondary scintillation solute is dimethyl POPOP. The variety of appreciably different surfactants is composed of isooctylphenol-polyethoxyethanol and sodium dihexyl sulphosuccinate [fr

  12. Synthesis of magnetic nanoparticles as a draw solute in forward osmosis membrane process for the treatment of radioactive liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Heeman; Lee, Kune Woo; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    These wastes contain about 0.3 ∼ 0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40 ∼ 90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. As an emerging technology forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination because FO operates at low or no hydraulic pressures. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe{sub 3}O{sub 4} nanoparticles can be separated from water by an external magnet field easily. If Fe{sub 3}O{sub 4} nanoparticles are coated with highly soluble organic substances, thus they can be used as a draw solute by concurrently generating high osmotic pressure and easy separation. The carboxylated polyglycerol coated Fe{sub 3}O{sub 4} nanoparticles have been successfully synthesized. The nanoparticles were about 50 nm in diameter and showed the good colloidal stability in aqueous solution. The osmolality and osmotic pressure were enough high to be used as a draw solute in FO. For the future work, we will investigate the performance of our magnetic draw solute in FO to remove boron in the simulated liquid waste.

  13. Synthesis of magnetic nanoparticles as a draw solute in forward osmosis membrane process for the treatment of radioactive liquid waste

    International Nuclear Information System (INIS)

    Yang, Heeman; Lee, Kune Woo; Moon, Jei Kwon

    2013-01-01

    These wastes contain about 0.3 ∼ 0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40 ∼ 90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. As an emerging technology forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination because FO operates at low or no hydraulic pressures. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe 3 O 4 nanoparticles can be separated from water by an external magnet field easily. If Fe 3 O 4 nanoparticles are coated with highly soluble organic substances, thus they can be used as a draw solute by concurrently generating high osmotic pressure and easy separation. The carboxylated polyglycerol coated Fe 3 O 4 nanoparticles have been successfully synthesized. The nanoparticles were about 50 nm in diameter and showed the good colloidal stability in aqueous solution. The osmolality and osmotic pressure were enough high to be used as a draw solute in FO. For the future work, we will investigate the performance of our magnetic draw solute in FO to remove boron in the simulated liquid waste

  14. Recovery of copper and cyanide from waste cyanide solutions using emulsion liquid membrane with LIX 7950 as the carrier.

    Science.gov (United States)

    Xie, Feng; Wang, Wei

    2017-08-01

    The feasibility of using emulsion liquid membranes (ELMs) with the guanidine extractant LIX 7950 as the mobile carrier for detoxifying copper-containing waste cyanide solutions has been determined. Relatively stable ELMs can be maintained under suitable stirring speed during mixing ELMs and the external solution. Effective extraction of copper cyanides by ELMs only occurs at pH below 11. High copper concentration in the external phase and high volume ratio of the external phase to ELMs result in high transport rates of copper and cyanide. High molar ratio of cyanide to copper tends to suppress copper extraction. The presence of thiocyanate ion significantly depresses the transport of copper and cyanide through the membrane while the thiosulfate ion produces less impact on copper removal by ELMs. Zinc and nickel cyanides can also be effectively extracted by ELMs. More than 90% copper and cyanide can be effectively removed from alkaline cyanide solutions by ELMs under suitable experimental conditions, indicating the effectiveness of using the designed ELM for recovering copper and cyanide from waste cyanide solutions.

  15. 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.)

  16. Highly water soluble nanoparticles as a draw solute in forward osmosis for the treatment of radioactive liquid waste

    International Nuclear Information System (INIS)

    Yang, Heeman; Choi, Hye Min; Jang, Sungchan; Seo, Bumkyoung; Lee, Kune Woo; Moon, Jei Kwon

    2014-01-01

    . In this study, we introduced highly water-soluble hyperbranched caroboxylated polyglycerol-coated magnetic nanoparticles (CPG-MNPs). It is known that the highly branched, globular architecture of PG significantly increase solubility compared to linear polymer and they are eco-friendly. The CPG-MNPs showed no aggregate of particles in water even after placing external magnet, and exhibited a high water flux in FO process. The CPG-MNPs are, therefore, potentially useful as a draw solute in FO processes. The operation of nuclear pressurized water reactors (PWRs) results in numerous radioactive waste streams which vary in radioactivity content. Most PWR stations have experienced leakages of boric acid into liquid radioactive waste systems. These wastes contain about 0.3∼0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40∼90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. Forward osmosis (FO), a low energy technique based on membrane technologies, has recently garnered attention for its utility in wastewater treatment and desalination applications. In the FO process, water flows across a semi-permeable membrane from a solution with a low osmotic pressure (the feed solution) to a solution with a high osmotic pressure (the draw solution). The driving force in FO processes is provided by the osmotic gradient between the two solutions. Low energy costs and low degrees of membrane fouling are two of the advantages conveyed by FO processes over other processes, such as reverse osmosis processes that rely on a hydraulic pressure driving force. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe3O4 nanoparticles can be separated from water by an external magnet field

  17. Highly water soluble nanoparticles as a draw solute in forward osmosis for the treatment of radioactive liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Heeman; Choi, Hye Min; Jang, Sungchan; Seo, Bumkyoung; Lee, Kune Woo; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    . In this study, we introduced highly water-soluble hyperbranched caroboxylated polyglycerol-coated magnetic nanoparticles (CPG-MNPs). It is known that the highly branched, globular architecture of PG significantly increase solubility compared to linear polymer and they are eco-friendly. The CPG-MNPs showed no aggregate of particles in water even after placing external magnet, and exhibited a high water flux in FO process. The CPG-MNPs are, therefore, potentially useful as a draw solute in FO processes. The operation of nuclear pressurized water reactors (PWRs) results in numerous radioactive waste streams which vary in radioactivity content. Most PWR stations have experienced leakages of boric acid into liquid radioactive waste systems. These wastes contain about 0.3∼0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40∼90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. Forward osmosis (FO), a low energy technique based on membrane technologies, has recently garnered attention for its utility in wastewater treatment and desalination applications. In the FO process, water flows across a semi-permeable membrane from a solution with a low osmotic pressure (the feed solution) to a solution with a high osmotic pressure (the draw solution). The driving force in FO processes is provided by the osmotic gradient between the two solutions. Low energy costs and low degrees of membrane fouling are two of the advantages conveyed by FO processes over other processes, such as reverse osmosis processes that rely on a hydraulic pressure driving force. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe3O4 nanoparticles can be separated from water by an external magnet field

  18. 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.)

  19. Liquid scintillation solution

    International Nuclear Information System (INIS)

    Long, E.C.

    1977-01-01

    A liquid scintillation solution is described which includes (1) a scintillation solvent (toluene and xylene), (2) a primary scintillation solute (PPO and Butyl PBD), (3) a secondary scintillation solute (POPOP and Dimethyl POPOP), (4) a plurality of substantially different surfactants and (5) a filter dissolving and/or transparentizing agent. 8 claims

  20. Ecological solution of the problem of handling liquid radioactive wastes - Lr (by the example of Flue SSC RF RIA R)

    International Nuclear Information System (INIS)

    Polyakov, V.I.; Bukvich, B.A.

    2006-01-01

    A sharp reduction of nuclear waste amounts is possible if their elements are considered as source material of atomic complexes - SMAC. The prospect of their possible salvaging will require technological changes and ensuring safety of storage of the material till the need arises. Long experience in deep liquid radioactive waste disposal and accounting, calculations, and motivations demonstrate that a corresponding choice of geological formations makes it possible to abandon liquid radioactive waste solidification and ensure their isolation from environment when the most rigid radiation safety requirements are fulfilled. (author)

  1. 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.

  2. Liquid scintillation solution

    International Nuclear Information System (INIS)

    Long, E.C.

    1976-01-01

    The invention deals with a liquid scintillation solution which contains 1) a scintillation solvent (toluol), 2) a primary scintillation solute (PPO), 3) a secondary scintillation solute (dimethyl POPOP), 4) several surfactants (iso-octyl-phenol polyethoxy-ethanol and sodium di-hexyl sulfosuccinate) essentially different from one another and 5) a filter resolution and/or transparent-making agent (cyclic ether, especially tetrahydrofuran). (HP) [de

  3. Radioactive waste management solutions

    International Nuclear Information System (INIS)

    Siemann, Michael

    2015-01-01

    One of the more frequent questions that arise when discussing nuclear energy's potential contribution to mitigating climate change concerns that of how to manage radioactive waste. Radioactive waste is produced through nuclear power generation, but also - although to a significantly lesser extent - in a variety of other sectors including medicine, agriculture, research, industry and education. The amount, type and physical form of radioactive waste varies considerably. Some forms of radioactive waste, for example, need only be stored for a relatively short period while their radioactivity naturally decays to safe levels. Others remain radioactive for hundreds or even hundreds of thousands of years. Public concerns surrounding radioactive waste are largely related to long-lived high-level radioactive waste. Countries around the world with existing nuclear programmes are developing longer-term plans for final disposal of such waste, with an international consensus developing that the geological disposal of high-level waste (HLW) is the most technically feasible and safe solution. This article provides a brief overview of the different forms of radioactive waste, examines storage and disposal solutions, and briefly explores fuel recycling and stakeholder involvement in radioactive waste management decision making

  4. 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.)

  5. 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)

  6. Liquid waste processing device

    International Nuclear Information System (INIS)

    Matsumoto, Kaname; Obe, Etsuji; Wakamatsu, Toshifumi.

    1989-01-01

    In a liquid waste processing device for processing living water wastes discharged from nuclear power plant facilities through a filtration vessel and a sampling vessel, a filtration layer disposed in the filtration vessel is divided into a plurality of layers along planes vertical to the direction of flow and the size of the filter material for each of the divided layers is made finer toward the downstream. Further, the thickness of the filtration material in each of the divided layers is also reduced toward the downstream. The filter material is packed such that the porosity in each of the divided layers is substantially identical. Further, the filtration material is packed in a mesh-like bag partitioned into a desired size and laid with no gaps to the planes vertical to the direction of the flow. Thus, liquid wastes such as living water wastes can be processed easily and simply so as to satisfy circumstantial criteria without giving undesired effects on the separation performance and life time and with easy replacement of filter. (T.M.)

  7. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.

  8. Spray drying of liquid radioactive wastes

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  9. Radioactive liquid waste processing method

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Liquid secondary waste: Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-31

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  11. Waste processing of chemical cleaning solutions

    International Nuclear Information System (INIS)

    Peters, G.A.

    1991-01-01

    This paper reports on chemical cleaning solutions containing high concentrations of organic chelating wastes that are difficult to reduce in volume using existing technology. Current methods for evaporating low-level radiative waste solutions often use high maintenance evaporators that can be costly and inefficient. The heat transfer surfaces of these evaporators are easily fouled, and their maintenance requires a significant labor investment. To address the volume reduction of spent, low-level radioactive, chelating-based chemical cleaning solutions, ECOSAFE Liquid Volume Reduction System (LVRS) has been developed. The LVRS is based on submerged combustion evaporator technology that was modified for treatment of low-level radiative liquid wastes. This system was developed in 1988 and was used to process 180,000 gallons of waste at Oconee Nuclear Station

  12. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting.

    Science.gov (United States)

    Chiang, Po-Neng; Tong, Ou-Yang; Chiou, Chyow-San; Lin, Yu-An; Wang, Ming-Kuang; Liu, Cheng-Chung

    2016-01-15

    A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg(-1) in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L(-1) DOC solution with a of pH 2.0 at 25°C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH4(+)-N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively. Copyright © 2015 Elsevier B.V. All rights reserved.

  13. 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

  14. 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.)

  15. Radiochromic liquid solution

    International Nuclear Information System (INIS)

    Noakes, J.E.; Culp, R.A.

    1983-01-01

    A radiochromic solution which is sensitive to small dosages of ionizing and ultraviolet radiation is described. It consists of a solution of a leucocyanide dye in a clear polar solvent with enough organic acid added to make the solution at least slightly acidic and responds to radiation by permanently changing color. Up to one half of the solution by weight can be replaced by a second solution of an aromatic solvent and an organic fluor. Another modification of the invention is a solution of a leucocyanide dye in a clear polar solvent having an aromatic group, an organic fluor, and enough organic acid to make the solution at least slightly acidic. (author)

  16. Radioactive liquid waste processing method

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. Electrical processes for liquid waste treatment

    International Nuclear Information System (INIS)

    Turner, A.D.; Bridger, N.J.; Junkison, A.R.; Pottinger, J.S.

    1987-08-01

    This report describes the development of electrical techniques for the treatment of liquid waste streams. Part I is concerned with solid/liquid separation and the demonstration of the electrokinetic thickening of flocs at inorganic membranes suitable for intermediate-level wastes and electrochemical cleaning of stainless steel microfilters and graphite ultrafilters. Part II describes work on the development of electrochemical ion exchange, particularly the use of inorganic absorption media and polarity reversal to enhance system selectivity. Work on the adsorption and desorption of plutonium in acid nitrate solution at various electrode materials is also included. (author)

  18. Method of solidifying radioactive liquid wastes

    International Nuclear Information System (INIS)

    Uetake, Naoto; Kawamura, Fumio; Kikuchi, Makoto; Fukazawa, Tetsuo.

    1983-01-01

    Purpose: To enable to confine the volatiling ingredients such as cesium in liquid wastes safely in glass solidification products while suppressing the volatilization thereof. Method: Acid salt of tetravalent metal such as titanium phosphate has an intense selective adsorption property to cesium. So liquid wastes stored in a high level liquid wastes tank is mixed with titanium phosphate gels stored in an adsorbent tank, then supplied to a mixer and mixed with a sodium silicate solution stored in a sodium silicate storage tank and boric acid stored in an additive tank, into gel-like state. The gel-like material thus formed is supplied to a drier. After being dried at a temperature of 200sup(o)C - 300sup(o)C, the material is melted under heating at a temperature of 1000sup(o)C - 1100sup(o)C, and then cooled to solidify. (Horiuchi, T.)

  19. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting

    International Nuclear Information System (INIS)

    Chiang, Po-Neng; Tong, Ou-Yang; Chiou, Chyow-San; Lin, Yu-An; Wang, Ming-Kuang; Liu, Cheng-Chung

    2016-01-01

    Highlights: • Nitrogen, phosphorus, and potassium contents in soil are substantially increased after the DOC washing. • The removal of Zn is dominated by proton replacement at pH 2.0, rather than by complexation with DOC. • The removal of Zn is dominated by DOC complexation between pH 3.0 and pH 5.0. - Abstract: A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg −1 in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L −1 DOC solution with a of pH 2.0 at 25 °C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH 4 + -N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively.

  20. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Po-Neng [Experimental Forest, National Taiwan University, Chushan, Nantou County, 55750, Taiwan (China); Tong, Ou-Yang [Department of Environment Engineering, College of the Environment and Ecology, and The Key Laboratory of the Ministry of Education for Coastal and Wetland Ecosystem, Xiamen University, Xiamen (China); Chiou, Chyow-San; Lin, Yu-An [Department of Environmental Engineering, National Ilan University, Ilan 26047, Taiwan (China); Wang, Ming-Kuang [Department of Animal Science, National Ilan University, Ilan 26047, Taiwan (China); Liu, Cheng-Chung, E-mail: ccliu@niu.edu.tw [Department of Agricultural Chemistry, National Taiwan University, Taipei 10617, Taiwan (China)

    2016-01-15

    Highlights: • Nitrogen, phosphorus, and potassium contents in soil are substantially increased after the DOC washing. • The removal of Zn is dominated by proton replacement at pH 2.0, rather than by complexation with DOC. • The removal of Zn is dominated by DOC complexation between pH 3.0 and pH 5.0. - Abstract: A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg{sup −1} in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L{sup −1} DOC solution with a of pH 2.0 at 25 °C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH{sub 4}{sup +}-N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively.

  1. Solid and liquid radioactive wastes

    International Nuclear Information System (INIS)

    Cluchet, J.; Desroches, J.

    1977-01-01

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

  2. Nuclear waste solutions

    Science.gov (United States)

    Walker, Darrel D.; Ebra, Martha A.

    1987-01-01

    High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  3. Solutions for Waste Management

    International Nuclear Information System (INIS)

    2013-01-01

    To safely and securely dispose of highlevel and long-lived radioactive waste, this material needs to be stored for a period of time that is very long compared to our everyday experience. Underground disposal facilities need to be designed and constructed in suitable geological conditions that can be confidently demonstrated to contain and isolate the hazardous waste from our environment for hundreds of thousands of years. Over this period of time, during which the safety of an underground waste repository system must be assured, the waste's radioactivity will decay to a level that cannot pose a danger to people or the environment. The archaeological record can help in visualizing such a long period of time. Climates change, oceans rise and vanish, and species evolve in the course of a one hundred millennia. Rocks bear witness to all of these changes. Geologists in their search for safe repositories for the long-term disposal of high level radioactive waste have identified rock formations that have proven stable for millions of years. These geological formations are expected to remain stable for millions of years and can serve as host formations for waste repositories.

  4. Treatment of low alpha activity liquid wastes

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Radioactive liquid wastes processing device

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  6. Reuse of hydroponic waste solution.

    Science.gov (United States)

    Kumar, Ramasamy Rajesh; Cho, Jae Young

    2014-01-01

    Attaining sustainable agriculture is a key goal in many parts of the world. The increased environmental awareness and the ongoing attempts to execute agricultural practices that are economically feasible and environmentally safe promote the use of hydroponic cultivation. Hydroponics is a technology for growing plants in nutrient solutions with or without the use of artificial medium to provide mechanical support. Major problems for hydroponic cultivation are higher operational cost and the causing of pollution due to discharge of waste nutrient solution. The nutrient effluent released into the environment can have negative impacts on the surrounding ecosystems as well as the potential to contaminate the groundwater utilized by humans for drinking purposes. The reuse of non-recycled, nutrient-rich hydroponic waste solution for growing plants in greenhouses is the possible way to control environmental pollution. Many researchers have successfully grown several plant species in hydroponic waste solution with high yield. Hence, this review addresses the problems associated with the release of hydroponic waste solution into the environment and possible reuse of hydroponic waste solution as an alternative resource for agriculture development and to control environmental pollution.

  7. New Approaches to Cleaning Liquid Radioactive Waste

    Directory of Open Access Journals (Sweden)

    Zabulonov, Yu.L.

    2015-05-01

    Full Text Available The industrial cleaning methods of liquid radioactive waste and technologically contaminated solutions, which contain heavy metals and radionuclides, are considered. It is shown that in the case when heavy metal ions exclusively exist in ionic form, the cleaning method with highest efficiency is electrodialysis. In the case when components, which must be removed, are in ionic and colloidal forms at the same time, the previous destruction of colloidal and organic matter (method of hydrodynamic cavitation, lowtemperature plasma etc is necessary. The developed «PTANK» method enables an effective purification of multicomponent metalcontaining man-made solutions, which contain additionally organic substances and complexes. Development of advanced membrane technologies, creation of complex recycling schemes and their synergistic combination will provide an opportunity to achieve deep cleaning of technologically contaminated solutions and minimize the amount of secondary wastes.

  8. 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.)

  9. 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.)

  10. Radioactive liquid waste processing device

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Reaction of H atoms with chelators in highly basic solution: H2 production in high level liquid waste simulants

    International Nuclear Information System (INIS)

    Barnabas, F.; Cerny, E.; Jonah, C.D.; Meisel, D.; Sauer, M.C. Jr.

    1995-01-01

    The rate constants for hydrogen abstraction by H from ethylene-diamine tetracetic acid (EDTA), N-(2-hydroxyethyl)-ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), glycolic acid and citric acid were measured at pH 13. The predominant product of these reactions is H 2 . The rate constants obtained are more than an order of magnitude larger than the literature rates that had been measured at pH 1. These measurements are of significance for understanding the radiolytic production of H 2 in nuclear waste storage tanks. (Author)

  12. 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.)

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. Liquid waste treatment system. Final report

    International Nuclear Information System (INIS)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of high-level liquid radioactive waste (HLW) at the West Valley Demonstration Project (WVDP) involved three distinct processing operations: decontamination of liquid HLW in the Supernatant Treatment System (STS); volume reduction of decontaminated liquid in the Liquid Waste Treatment System (LWTS); and encapsulation of resulting concentrates into an approved cement waste form in the Cement Solidification System (CSS). Together, these systems and operations made up the Integrated Radwaste Treatment System (IRTS)

  18. Study of liquids and solutions

    International Nuclear Information System (INIS)

    Bellissent-Funel, M.C.

    1994-01-01

    A critical review of what has been achieved on the structure of liquids and solutions and the capabilities and developments of neutron scattering in this domain, are presented. A great variety of simple to complex systems has been investigated with the aim of obtaining a full microscopic description of the structure. Selected examples demonstrate the neutron scattering determination of interaction potentials, intermolecular structures and partial structure factors of complex systems. The isotopic substitution method is illustrated by the application to the study of the solvation of ions in aqueous and non aqueous solutions. (author). 9 figs., 32 refs

  19. Selective separation of cesium from simulated high level liquid waste solution using 1,3-dioctyloxy calix[4]arene-benzo-crown-6

    International Nuclear Information System (INIS)

    Vikas Kumar; Sharma, J.N.; Hubli, R.C.

    2014-01-01

    The 25,27-di(octyloxy)calix[4]arenebenzocrown-6 (CBC) in 1,3-alternate conformation was synthesized indigenously starting from its intermediates in good yield and purity. The extraction studies of CBC were carried out by using two different phase modifiers namely isodecyl alcohol and ortho-nitrophenyl hexyl ether. Detailed investigations on the effect of various parameters like, concentration of phase modifiers, aqueous phase acidity, ligand concentration, nitrate ion concentration and effect of temperature on extraction of cesium have been carried out. The concentration of phase modifiers was optimized to be 30 % in n-dodecane to ensure optimum extraction of cesium. Stoichiometry of the extracted complex determined by slope analysis method reveals 1:1:1 molar ratio for CsNO 3 :CBC:HNO 3 . The extraction process was found to be exothermic as determined from the plot of log K ex versus 1/T. The solvent system with a composition 0.01 M CBC/30 % phase modifier/n-dodecane was found to be effective for selective separation of cesium from simulated high level liquid waste solution. (author)

  20. 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.)

  1. Electrochemical treatment of liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D.T. [Savannah River Technology Center, Aiken, SC (United States)

    1997-10-01

    Under this task, electrochemical treatment processes are being evaluated and developed for the destruction of organic compounds and nitrates/nitrites and the removal of other hazardous species from liquid wastes stored throughout the DOE complex. This technology targets the (1) destruction of nitrates, nitrites and organic compounds; (2) removal of radionuclides; and (3) removal of RCRA metals. The development program consists of five major tasks: (1) evaluation of electrochemical reactors for the destruction and removal of hazardous waste components, (2) development and validation of engineering process models, (3) radioactive laboratory-scale tests, (4) demonstration of the technology in an engineering-scale reactor, and (5) analysis and evaluation of test data. The development program team is comprised of individuals from national laboratories, academic institutions, and private industry. Possible benefits of this technology include: (1) improved radionuclide separation as a result of the removal of organic complexants, (2) reduction in the concentrations of hazardous and radioactive species in the waste (e.g., removal of nitrate, mercury, chromium, cadmium, {sup 99}Tc, and {sup 106}Ru), (3) reduction in the size of the off-gas handling equipment for the vitrification of low-level waste (LLW) by reducing the source of NO{sub x} emissions, (4) recovery of chemicals of value (e.g. sodium hydroxide), and (5) reduction in the volume of waste requiring disposal.

  2. Liquid Secondary Waste Grout Formulation and Waste Form Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-23

    This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption Kd (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.

  3. 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

  4. Electrochemical treatment of liquid wastes

    International Nuclear Information System (INIS)

    Hobbs, D.

    1996-01-01

    Electrochemical treatment processes are being evaluated and developed for the destruction of organic compounds and nitrates/nitrites and the removal of other hazardous species from liquid wastes stored throughout the DOE complex. This activity consists of five major tasks: (1) evaluation of different electrochemical reactors for the destruction and removal of hazardous waste components, (2) development and validation of engineering process models, (3) radioactive laboratory-scale tests, (4) demonstration of the technology in an engineering-scale size reactor, and (5) analysis and evaluation of testing data. The development program team is comprised of individuals from federal, academic, and private industry. Work is being carried out in DOE, academic, and private industrial laboratories

  5. Treatment of liquid radioactive waste: Precipitation

    International Nuclear Information System (INIS)

    Gompper, K.

    1982-01-01

    After introductory remarks about waste types to be treated, specific treatment methods are discussed and examples are given for treatment processes carried out with different types of liquid wastes from nuclear power plants, research centers and fuel reprocessing plants. (RW)

  6. Method of processing radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-21

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

  7. 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.)

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

    International Nuclear Information System (INIS)

    1984-01-01

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

  9. Scientific Solutions to Nuclear Waste Environmental Challenges

    International Nuclear Information System (INIS)

    Johnson, Bradley R.

    2014-01-01

    The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount

  10. 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

  11. Final treatment of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Svolik, S.

    2004-01-01

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

  12. Management of liquid radioactive wastes at PNRI

    International Nuclear Information System (INIS)

    Garcia, C.M.

    1994-10-01

    Liquid wastes accepted at PNRI waste management facility are generated by hospitals and research institutions from all over the country including those generated from the research laboratories within the PNRI. The operation of the Philippine TRIGA Research Reactor is also a potential source of liquid waste to be handled and managed by the facility in the future. This technical report is a result of the study of the present status and development of the management of liquid wastes at PNRI. (auth.). 8 refs.; 3 figs.; 4 tabs

  13. Liquid waste processing at Comanche Peak

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. Method of processing decontaminating liquid waste

    International Nuclear Information System (INIS)

    Kusaka, Ken-ichi

    1989-01-01

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

  15. Supported liquid inorganic membranes for nuclear waste separation

    Science.gov (United States)

    Bhave, Ramesh R; DeBusk, Melanie M; DelCul, Guillermo D; Delmau, Laetitia H; Narula, Chaitanya K

    2015-04-07

    A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

  16. 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)

  17. The immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1986-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a 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

  18. Study on cementation of simulated radioactive borated liquid wastes

    International Nuclear Information System (INIS)

    Sun Qina; Li Junfeng; Wang Jianlong

    2010-01-01

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

  19. Waste Treatment Plant Liquid Effluent Treatability Evaluation

    International Nuclear Information System (INIS)

    LUECK, K.J.

    2001-01-01

    Bechtel National, Inc. (BNI) provided a forecast of the radioactive, dangerous liquid effluents expected to be generated by the Waste Treatment Plant (WTP). The forecast represents the liquid effluents generated from the processing of 25 distinct batches of tank waste through the WTP. The WTP liquid effluents will be stored, treated, and disposed of in the Liquid Effluent Retention Facility (LERF) and the Effluent Treatment Facility (ETF). Fluor Hanford, Inc. (FH) evaluated the treatability of the WTP liquid effluents in the LERFIETF. The evaluation was conducted by comparing the forecast to the LERFIETF treatability envelope, which provides information on the items that determine if a liquid effluent is acceptable for receipt and treatment at the LERFIETF. The WTP liquid effluent forecast is outside the current LERFlETF treatability envelope. There are several concerns that must be addressed before the WTP liquid effluents can be accepted at the LERFIETF

  20. Membrane separation of ionic liquid solutions

    Science.gov (United States)

    Campos, Daniel; Feiring, Andrew Edward; Majumdar, Sudipto; Nemser, Stuart

    2015-09-01

    A membrane separation process using a highly fluorinated polymer membrane that selectively permeates water of an aqueous ionic liquid solution to provide dry ionic liquid. Preferably the polymer is a polymer that includes polymerized perfluoro-2,2-dimethyl-1,3-dioxole (PDD). The process is also capable of removing small molecular compounds such as organic solvents that can be present in the solution. This membrane separation process is suitable for drying the aqueous ionic liquid byproduct from precipitating solutions of biomass dissolved in ionic liquid, and is thus instrumental to providing usable lignocellulosic products for energy consumption and other industrial uses in an environmentally benign manner.

  1. Effectiveness of liquid radioactive waste purification by inorganic granulated sorbents

    International Nuclear Information System (INIS)

    Komarevskij, V.M.; Stepanets, O.V.; Sharygin, L.M.; Matveev, S.A.

    1995-01-01

    Study results on purification of simulative and real liquid radioactive wastes from fission products radionuclides and by inorganic corrosion-nature sorbents 'Thermoxide' are presented. Properties by sorption of cesium, strontium and cobalt are studied; results of experiments on purification of weakly-salted water solutions (waste waters, ships drainage tanks, showers and laundries) of the Beloyarsk NPP are presented. Sorbents source characteristics are determined. 4 refs., 2 figs., 3 tabs

  2. Studies on radioactive liquid waste treatment by reverse osmosis

    International Nuclear Information System (INIS)

    Koyama, Akio; Shimoura, Kazukuni; Tsutsui, Tenson

    1982-01-01

    Reverse osmosis is a simple process and has relatively high decontamination factor comparing to other processes used for the treatment of radioactive liquid waste. Furthermore the quantity of secondary waste of this process is small. In this study, test solution containing nine elements such as cesium, strontium, cobalt etc. in chloride forms are treated by reverse osmosis. Permeate rate decreases as the increase of osmotic pressure of feed solution and is expressed by linear equation. Decontamination factor of cations of univalency is more than ten, and about one tenth of that of bivalency. Decontamination factors of all the elements used in this experiment are approximately estimated using the solution-diffusion model. (author)

  3. Reduction of INTEC Analytical Radioactive Liquid Wastes

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  4. Photochemical oxidation: A solution for the mixed waste dilemma

    Energy Technology Data Exchange (ETDEWEB)

    Prellberg, J.W.; Thornton, L.M.; Cheuvront, D.A. [Vulcan Peroxidation Systems, Inc., Tucson, AZ (United States)] [and others

    1995-12-31

    Numerous technologies are available to remove organic contamination from water or wastewater. A variety of techniques also exist that are used to neutralize radioactive waste. However, few technologies can satisfactorily address the treatment of mixed organic/radioactive waste without creating unacceptable secondary waste products or resulting in extremely high treatment costs. An innovative solution to the mixed waste problem is on-site photochemical oxidation. Liquid-phase photochemical oxidation has a long- standing history of successful application to the destruction of organic compounds. By using photochemical oxidation, the organic contaminants are destroyed on-site leaving the water, with radionuclides, that can be reused or disposed of as appropriate. This technology offers advantages that include zero air emissions, no solid or liquid waste formation, and relatively low treatment cost. Discussion of the photochemical process will be described, and several case histories from recent design testing, including cost analyses for the resulting full-scale installations, will be presented as examples.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  6. Recovery of Mercury From Contaminated Liquid Wastes

    International Nuclear Information System (INIS)

    1998-01-01

    The Base Contract program emphasized the manufacture and testing of superior sorbents for mercury removal, testing of the sorption process at a DOE site, and determination of the regeneration conditions in the laboratory. During this project, ADA Technologies, Inc. demonstrated the following key elements of a successful regenerable mercury sorption process: (1) sorbents that have a high capacity for dissolved, ionic mercury; (2) removal of ionic mercury at greater than 99% efficiency; and (3) thermal regeneration of the spent sorbent. ADA's process is based on the highly efficient and selective sorption of mercury by noble metals. Contaminated liquid flows through two packed columns that contain microporous sorbent particles on which a noble metal has been finely dispersed. A third column is held in reserve. When the sorbent is loaded with mercury to the point of breakthrough at the outlet of the second column, the first column is taken off-line and the flow of contaminated liquid is switched to the second and third columns. The spent column is regenerated by heating. A small flow of purge gas carries the desorbed mercury to a capture unit where the liquid mercury is recovered. Laboratory-scale tests with mercuric chloride solutions demonstrated the sorbents' ability to remove mercury from contaminated wastewater. Isotherms on surrogate wastes from DOE's Y-12 Plant in Oak Ridge, Tennessee showed greater than 99.9% mercury removal. Laboratory- and pilot-scale tests on actual Y-12 Plant wastes were also successful. Mercury concentrations were reduced to less than 1 ppt from a starting concentration of 1,000 ppt. The treatment objective was 50 ppt. The sorption unit showed 10 ppt discharge after six months. Laboratory-scale tests demonstrated the feasibility of sorbent regeneration. Results show that sorption behavior is not affected after four cycles

  7. 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.)

  8. 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

  9. Transport of Liquid Phase Organic Solutes in Liquid Crystalline Membranes

    OpenAIRE

    Han, Sangil

    2010-01-01

    Porous cellulose nitrate membranes were impregnated with 8CB and PCH5 LCs (liquid crystals) and separations of solutes dissolved in aqueous phases were performed while monitoring solute concentration via UV-VIS spectrometry. The diffusing organic solutes, which consist of one aromatic ring and various functional groups, were selected to exclude molecular size effects on the diffusion and sorption. We studied the effects on solute transport of solute intra-molecular hydrogen bonding and so...

  10. 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)

  11. Method for storage of liquid radioactive waste

    International Nuclear Information System (INIS)

    Hesky, H.; Wunderer, A.

    1978-01-01

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

  12. 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

  13. Treatment of mixed radioactive liquid wastes at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Chamberlain, D.B.; Conner, C.

    1994-01-01

    Aqueous mixed waste at Argonne National Laboratory (ANL) is traditionally generated in small volumes with a wide variety of compositions. A cooperative effort at ANL between Waste Management (WM) and the Chemical Technology Division (CMT) was established, to develop, install, and implement a robust treatment operation to handle the majority of such wastes. For this treatment, toxic metals in mixed-waste solutions are precipitated in a semiautomated system using Ca(OH) 2 and, for some metals, Na 2 S additions. This step is followed by filtration to remove the precipitated solids. A filtration skid was built that contains several filter types which can be used, as appropriate, for a variety of suspended solids. When supernatant liquid is separated from the toxic-metal solids by decantation and filtration, it will be a low-level waste (LLW) rather than a mixed waste. After passing a Toxicity Characteristic Leaching Procedure (TCLP) test, the solids may also be treated as LLW

  14. Treatment of liquid wastes from uranium hydrometallurgy

    International Nuclear Information System (INIS)

    Moraga G, J.C.

    1988-01-01

    Different treatments for low activity liquid wastes, generated by the hidromettalurgy of uranium ore are studied. A process of treatment was chosen which includes a neutralization with lime and limestone and a selective removal of Ra-226, through ion-exchange resins. A plant, with a capacity of treatment of 1 m 3 /h of liquid effluents was scoped. (author)

  15. Liquid-liquid interfacial tension of electrolyte solutions

    NARCIS (Netherlands)

    Bier, Markus; Zwanikken, J.W.; van Roij, R.H.H.G.

    2008-01-01

    It is theoretically shown that the excess liquid-liquid interfacial tension between two electrolyte solutions as a function of the ionic strength I behaves asymptotically as (-) for small I and as (±I) for large I. The former regime is dominated by the electrostatic potential due to an unequal

  16. Solid and Liquid Waste Drying Bag

    Science.gov (United States)

    Litwiller, Eric (Inventor); Hogan, John A. (Inventor); Fisher, John W. (Inventor)

    2009-01-01

    Method and system for processing waste from human activities, including solids, liquids and vapors. A fluid-impermeable bag, lined with a liquid-impermeable but vapor-permeable membrane, defining an inner bag, is provided. A vacuum force is provided to extract vapors so that the waste is moved toward a selected region in the inner bag, extracted vapors, including the waste vapors and vaporized portions of the waste liquids are transported across the membrane, and most or all of the solids remain within the liner. Extracted vapors are filtered, and sanitized components thereof are isolated and optionally stored. The solids remaining within the liner are optionally dried and isolated for ultimate disposal.

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

    International Nuclear Information System (INIS)

    Fueriova, A.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  19. Convective instabilities in liquid centrifugation for nuclear wastes separation

    Energy Technology Data Exchange (ETDEWEB)

    Camassa, R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The separation of fission products from liquid solutions using centrifugal forces may prove an effective alternative to chemical processing in cases where radioactive materials necessitate minimal mixed-waste products or when allowing access to sophisticated chemical processing is undesirable. This investigation is a part of the effort to establish the feasibility of using liquid centrifugation for nuclear waste separation in the Accelerator Driven Energy Production (ADEP) program. A number of fundatmental issues in liquid centrifugation with radioactive elements need to be addressed in order to validate the approach and provide design criteria for experimental liquid salt (LiF and BeF{sub 2}) centrifuge. The author concentrates on one such issue, the possible onset of convective instabilities which could inhibit separation.

  20. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration.

  1. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    International Nuclear Information System (INIS)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon

    2014-01-01

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration

  2. Radiation degradation of waste waters. Reverse phase-high performance liquid chromatography and multicomponent UV-VIS analysis of gamma-irradiated aqueous solutions of nitrobenzene Pt.1

    International Nuclear Information System (INIS)

    Kuruc, J.; Sahoo, M.K.; Locaj, J.; Hutta, M.

    1994-01-01

    Saturated aqueous solutions of nitrobenzene (in water, 0.1M nitric acid and 0.1M potassium hydroxide) were irradiated with 60 Co γ-rays in deaerated condition. Radiolytic products were analyzed using reverse phase-high performance liquid chromatography (RP-HPLC) and multicomponent UV-VIS spectrometry. With the aid of RP-HPLC retention times of the radiolytic products were found to be identical with those of isomeric nitrophenols, aminophenols and dinitrophenols. According to the primary information obtained from RP-HPLC and literature, we have chosen ten standards and eleven wavelengths for multicomponent UV-VIS analysis (linear multiparametric regression analysis) and the concentrations of nitrobenzene, nitrophenols, aminophenols and dinitrophenols in water, HNO 3 and KOH solutions were calculated. G-values (molecules/100 eV) of the radiolytic products and decomposition of nitrobenzene in aqueous solutions G(-nitrobenzene) were calculated from the dependence of their concentrations with dose. Ph has relatively little influence on the decrease of concentration of nitrobenzene, but has strong influence on the product composition. (author) 7 refs.; 5 figs.; 5 tabs

  3. Aqueous solutions of ionic liquids: microscopic assembly

    NARCIS (Netherlands)

    Vicent-Luna, J.M.; Dubbeldam, D.; Gómez-Álvarez, P.; Calero, S.

    2016-01-01

    Aqueous solutions of ionic liquids are of special interest, due to the distinctive properties of ionic liquids, in particular, their amphiphilic character. A better understanding of the structure-property relationships of such systems is hence desirable. One of the crucial molecular-level

  4. 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)

  5. Treatment of liquid radioactive waste: Evaporation

    International Nuclear Information System (INIS)

    Pfeiffer, R.

    1982-01-01

    About 10.000 m 3 of low active liquid waste (LLW) arise in the Nuclear Research Center Karlsruhe. Chemical contents of this liquid waste are generally not declared. Resulting from experiments carried out in the Center during the early sixties, the evaporator facility was built in 1968 for decontamination of LLW. The evaporators use vapor compression and concentrate recirculation in the evaporator sump by pumps. Since 1971 the medium active liquid waste (MLW) from the Karlsruhe Reprocessing Plant (WAK) was decontaminated in this evaporator facility, too. By this time the amount of low liquid waste (LLW) had been decontaminated without mentionable interruptions. Afterwards a lot of interruptions of operations occurred, mainly due to leakages of pumps, valves and pipes. There was also a very high radiation level for the operating personnel. As a consequence of this experience a new evaporator facility for decontamination of medium active liquid waste was built in 1974. This facility started operation in 1976. The evaporator has natural circulation and is heated by steam through a heat exchanger. (orig./RW)

  6. INEEL Radioactive Liquid Waste Reduction Program

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  7. 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

  8. Method of decontamination for uranium oxide particles floating in liquid waste

    International Nuclear Information System (INIS)

    Terakado, Tsutomu; Ebara, Tsuneo; Sato, Kuniaki.

    1981-01-01

    Purpose: To rapidly treat liquid waste containing uranium oxide particles floating in it and to enable substantially complete decontamination. Method: An iron salt such as ferrous sulfate or the like is added to liquid waste with floating uranium oxide particles, an alkaline solution such as caustic soda or the like is then added to the liquid waste while feeding compressed air at 0.1 to 0.02 l/sec. per ton of liquid waste, and the pH of the liquid waste is made to from 6.5 to 7.5. Thereafter, the feed of compressed air is stopped, the liquid waste is allowed to stand, and is then filtered. (Aizawa, K.)

  9. Liquid Radioactive Wastes Treatment: A Review

    Directory of Open Access Journals (Sweden)

    Yung-Tse Hung

    2011-05-01

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

  10. Low level radioactive liquid waste decontamination by electrochemical way

    International Nuclear Information System (INIS)

    Tronche, E.

    1994-10-01

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

  11. 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.

  12. 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.

  13. 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)

  14. Conditioning of radioactive waste solutions by cementation

    International Nuclear Information System (INIS)

    Vejmelka, P.; Rudolph, G.; Kluger, W.; Koester, R.

    1992-02-01

    For the cementation of the low and intermediate level evaporator concentrates resulting from the reprocessing of spent fuel numerous experiments were performed to optimize the waste form composition and to characterize the final waste form. Concerning the cementation process, properties of the waste/cement suspension were investigated. These investigations include the dependence of viscosity, bleeding, setting time and hydration heat from the waste cement slurry composition. For the characterization of the waste forms, the mechanical, thermal and chemical stability were determined. For special cases detailed investigations were performed to determine the activity release from waste packages under defined mechanical and thermal stresses. The investigations of the interaction of the waste forms with aqueous solutions include the determination of the Cs/Sr release, the corrosion resistance and the release of actinides. The Cs/Sr release was determined in dependence of the cement type, additives, setting time and sample size. (orig./DG) [de

  15. 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

  16. Liquid low level waste management expert system

    International Nuclear Information System (INIS)

    Ferrada, J.J.; Abraham, T.J.; Jackson, J.R.

    1991-01-01

    An expert system has been developed as part of a new initiative for the Oak Ridge National Laboratory (ORNL) systems analysis program. This expert system will aid in prioritizing radioactive waste streams for treatment and disposal by evaluating the severity and treatability of the problem, as well as the final waste form. The objectives of the expert system development included: (1) collecting information on process treatment technologies for liquid low-level waste (LLLW) that can be incorporated in the knowledge base of the expert system, and (2) producing a prototype that suggests processes and disposal technologies for the ORNL LLLW system. 4 refs., 9 figs

  17. Liquid-liquid interfacial tension of electrolyte solutions

    OpenAIRE

    Bier, Markus; Zwanikken, Jos; van Roij, Rene

    2008-01-01

    It is theoretically shown that the excess liquid-liquid interfacial tension between two electrolyte solutions as a function of the ionic strength I behaves asymptotically as O(- I^0.5) for small I and as O(+- I) for large I. The former regime is dominated by the electrostatic potential due to an unequal partitioning of ions between the two liquids whereas the latter regime is related to a finite interfacial thickness. The crossover between the two asymptotic regimes depends sensitively on mat...

  18. Cementation of liquid radioactive waste

    International Nuclear Information System (INIS)

    Efremenkov, V.

    2004-01-01

    The cementation methods for immobilisation of radioactive wastes are discussed in terms of methodology, chemistry and properties of the different types of cements as well as the worldwide experience in this field. Two facilities for cementation - DEWA and MOWA - are described in details

  19. 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

  20. Bituminization of liquid radioactive waste. Part 3

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  1. 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)

  2. Wow Technology’s innovative radioactive liquid waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Marin, A.

    2015-07-01

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

  3. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. 40 CFR 761.269 - Sampling liquid PCB remediation waste.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Sampling liquid PCB remediation waste..., AND USE PROHIBITIONS Cleanup Site Characterization Sampling for PCB Remediation Waste in Accordance with § 761.61(a)(2) § 761.269 Sampling liquid PCB remediation waste. (a) If the liquid is single phase...

  5. Cs separation from nitric acid solutions of radioactive waste

    International Nuclear Information System (INIS)

    Heckmann, K.; Pieronczyk, W.; Strnad, J.; Feldmaier, F.

    1989-01-01

    It was the objective of this study to selectively separate active caesium (Cs-134 and Cs-137) from acid radioactive waste solutions (especially MAW and HAWC). The following 'strategy' was designed for a separation process: synthesis of reagents which are acid-resistant and selective for caesium; precipitation of Cs + and separation of the precipitates by filtration or centrifugation or precipitation of Cs + and separation of the precipitates by flotation; caesium separation by liquid-liquid extraction. As precipitating agents, sodium tetraphenylborate (kalignost) and several of its fluorine derivatives were examined. (orig./RB) [de

  6. Liquid wastes concentrating and solidifying device

    International Nuclear Information System (INIS)

    Kamiyoshi, Hideki; Ninokata, Yoshihide.

    1985-01-01

    Purpose: To provide a device for concentrating to solidify radioactive liquid wastes at large solidifying speed and with high decontaminating coefficient, without requirement for automatic control. Constitution: An asphalt solidifying device is disposed below a centrifugal thin film drier, and powder resulted from the drier is directly solidified with asphalt by utilizing the rotation of the drier for the mixing operation in the asphalt vessel. If abnormality should occur in the operation of the drier, resulting liquid wastes can be received and solidified in the asphalt vessel. The liquid wastes are heated to dry in a vessel main body having the heating surface at the circumferential surface. The vessel main body provided with a nozzle for supplying liquid to be treated disposed slantwise at the upper portion of the heating face, scrapers which rotate and slidingly contact the heating face and nozzles which jet out chemicals to the heating face behind the scrapers. Below the vessel main body, are disposed a funnel-like hopper for receiving falling scales, rotary vanes, and the likes by which the scales are introduced into the asphalt solidifying vessel. (Moriyama, K.)

  7. Combustion chamber for solid and liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Vcelak, L.; Kocica, J.; Trnobransky, K.; Hrubes, J. (VSCHT, Prague (Czechoslovakia))

    1989-04-01

    Describes combustion chamber incorporated in a new boiler manufactured by Elitex of Kdyne to burn waste products and occasionally liquid and solid waste from neighboring industries. It can handle all kinds of solids (paper, plastics, textiles, rubber, household waste) and liquids (volatile and non-volatile, zinc, chromium, etc.) and uses coal as a fuel additive. Its heat output is 3 MW, it can burn 1220 kg/h of coal (without waste, calorific value 11.76 MJ/kg) or 500 kg/h of coal (as fuel additive, calorific value 11.76 MJ/kg) or 285 kg/h of solid waste (calorific value 20.8 MJ/kg). Efficiency is 75%, capacity is 103 m{sup 3} and flame temperature is 1,310 C. Individual components are designed for manufacture in small engineering workshops with basic equipment. A disk absorber with alkaline filling is fitted for removal of harmful substances arising when PVC or tires are combusted.

  8. 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)

  9. 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)

  10. Liquid waste processing from TRIGA spent fuel storage pits

    International Nuclear Information System (INIS)

    Buchtela, Karl

    1988-01-01

    At the Atominstitute of the Austrian Universities and also at other facilities running TRIGA reactors, storage pits for spent fuel elements are installed. During the last revision procedure, the reactor group of the Atominstitute decided to refill the storage pits and to get rid of any contaminated storage pit water. The liquid radioactive waste had been pumped to polyethylene vessels for intermediate storage before decontamination and release. The activity concentration of the storage pit water at the Aominstitute after a storage period of several years was about 40 kBq/l, the total amount of liquid in the storage pits was about 0.25 m 3 . It was attempted to find a simple and inexpensive method to remove especially the radioactive Cesium from the waste solution. Different methods for decontamination like distillation, precipitation and ion exchange are discussed

  11. Corrosion of steel tanks in liquid nuclear wastes

    International Nuclear Information System (INIS)

    Carranza, Ricardo M.; Giordano, Celia M.; Saenz, Eduardo

    2005-01-01

    The objective of this work is to understand how solution chemistry would impact on the corrosion of waste storage steel tanks at the Hanford Site. Future tank waste operations are expected to process wastes that are more dilute with respect to some current corrosion inhibiting waste constituents. Assessment of corrosion damage and of the influence of exposure time and electrolyte composition, using simulated (non-radioactive) wastes, of the double-shell tank wall carbon steel alloys is being conducted in a statistically designed long-term immersion experiment. Corrosion rates at different times of immersion were determined using both weight-loss determinations and electrochemical impedance spectroscopy measurements. Localized corrosion susceptibility was assessed using short-term cyclic potentiodynamic polarization curves. The results presented in this paper correspond to electrochemical and weight-loss measurements of the immersed coupons during the first year of immersion from a two year immersion plan. A good correlation was obtained between electrochemical measurements, weight-loss determinations and visual observations. Very low general corrosion rates ( -1 ) were estimated using EIS measurements, indicating that general corrosion rate of the steel in contact with liquid wastes would no be a cause of tank failure even for these out-of-chemistry limit wastes. (author) [es

  12. 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.)

  13. Radioactive wastes. The groundwork of current solutions

    International Nuclear Information System (INIS)

    Grevoz, A.; Boullis, B.; Devezeaux de Lavergne, J.G.; Butez, M.; Bordier, G.; Vitart, X.; Hablot, I.; Chastagnet, F.

    2005-01-01

    Today the groundwork laid down by research has made processes available for the durable treatment and conditioning of all types of radioactive waste. This document illustrates the today situations in five presentations. Now standing as a national reference, the french inventory of radioactive waste, drawn up by ANDRA, has not only expanded to cover recoverable material but also features predictions of waste arisings for 2010 and 2020, including waste from the decommissioning of current installations. The current process used for spent fuel reprocessing allows extraction for recycling purpose, of uranium and plutonium, with very high recovery and purification rates. Advances in characterization and decontamination allow improvements in sorting and retrieval and conditioning to be considered for older wastes. The french National radioactive waste management agency (ANDRA) is already providing optimum industrial solutions for all short-lived, low and very low level waste on its Soulaines and Morvillers sites. For several decades, Areva has been reprocessing spent fuel and conditioning ultimate waste in its La Hague plants. (A.L.B.)

  14. 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

  15. Bituminization of liquid radioactive wastes. Part 1

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. Liquid waste processing from plutonium (III) oxalate precipitation

    International Nuclear Information System (INIS)

    Esteban, A.; Cassaniti, P.; Orosco, E.H.

    1990-01-01

    Plutonium (III) oxalate filtrates contain about 0.2M oxalic acid, 0.09M ascorbic acid, 0.05M hydrazine, 1M nitric acid and 20-100 mg/l of plutonium. The developed treatment of liquid wastes consist in two main steps: a) Distillation to reduce up to 10% of the initial volume and refluxing to destroy organic material. Then, the treated solution is suitable to adjust the plutonium at the tetravalent state by addition of hydrogen peroxide and the nitric molarity up to 8.6M. b) Recovery and purification of plutonium by anion exchange using two columns in series containing Dowex 1-X4 resin. With the proposed process, it is possible to transform 38 litres of filtrates with 40mg/l of Pu into 0.1 l of purified solution with 15-20g/l of Pu. This solution is suitable to be recycled in the Pu (III) oxalate precipitation process. This process has several potential advantages over similar liquid waste treatments. These include: 1) It does not increase the liquid volume. 2) It consumes only few reagents. 3) The operations involved are simple, requiring limited handling and they are feasible to automatization. 4) The Pu recovery factor is about 99%. (Author) [es

  17. 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.)

  18. Radioactive waste management turning options into solution

    International Nuclear Information System (INIS)

    Neubauer, J.

    2000-10-01

    Most of the statements from representatives of different countries and institutions focused on the status of high level radioactive waste management, including spent fuel repositories. Speakers dealing with such topics were representatives from countries applying nuclear power for electricity production. They all reported about there national programs on technical and safety aspects of radioactive waste management. The panel discussion extended to questions on political sensitivities and public acceptance; in this respect, interesting developments are taking place in Finland and Sweden. It is expected that Finland will operate a final repository for spent fuel in 10 - 15 years from now, followed close by Sweden. Other countries, however, face decisions by policy makers and elected officials to postpone dealing with waste disposal concerns. In this connection there is relevant experience in our country, too - even in the absence of spent fuel or other high level waste to be dealt with. During personal discussions with representatives of other countries not using nuclear power it was confirmed that there are similar or shared experiences. Development of publicly -accepted solutions to radioactive waste management remains an important issue. Independent of the amount or the activity of radioactive waste, the public at large remains skeptical despite the agreement among experts that disposal can be safe, technically feasible and environmentally sound. In countries not using nuclear power there are only small quantities of low and intermediate level radioactive waste. Therefore, international co-operation among such countries should be an option. There was common understanding by representatives from Norway, Italy and Austria that international co-operation should be developed for treatment and disposal of such waste. For the moment however it has to be accepted that, for political reasons, it is not possible. Forced to deal with the lack of near-term solutions, the

  19. Method of vitrificating fine-containing liquid waste

    International Nuclear Information System (INIS)

    Hagiwara, Minoru; Matsunaka, Kazuhisa.

    1989-01-01

    This invention concerns a vitrificating method of liquid wastes containing fines (metal powder discharged upon cutting fuel cans) used in a process for treating high level radioactive liquid wastes or a process for treating liquid wastes from nuclear power plants. Liquid wastes containing fines, slurries, etc. are filtered by a filter vessel comprising glass fibers. The fines are supplied as they are to a glass melting furnace placed in the vessel. Filterates formed upon filteration are mixed with other high level radioactive wastes and supplied together with starting glass material to the glass melting furnace. Since the fine-containing liquid wastes are processed separately from high radioactive liquid wastes, clogging of pipeways, etc. can be avoided, supply to the melting furnace is facilitated and the operation efficiency of the vitrification process can be improved. (I.N.)

  20. Boron removal in radioactive liquid waste by forward osmosis membrane

    Energy Technology Data Exchange (ETDEWEB)

    Doo Seong Hwang; Hei Min Choi; Kune Woo Lee; Jei Kwon Moon [KAERI, Daejeon (Korea, Republic of)

    2013-07-01

    This study investigated the treatment of boric acid contained in liquid radioactive waste using a forward osmosis membrane. The boron permeation through the membrane depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7 and increases with an increase of the osmotic driving force. The boron flux decreases slightly with the salt concentration, but is not heavily influenced by a low salt concentration. The boron flux increases linearly with the concentration of boron. No element except for boron was permeated through the FO membrane in the multi-component system. The maximum boron flux is obtained in an active layer facing a draw solution orientation of the CTA-ES membrane under conditions of less than pH 7 and high osmotic pressure. (authors)

  1. Ionic liquids behave as dilute electrolyte solutions

    Science.gov (United States)

    Gebbie, Matthew A.; Valtiner, Markus; Banquy, Xavier; Fox, Eric T.; Henderson, Wesley A.; Israelachvili, Jacob N.

    2013-01-01

    We combine direct surface force measurements with thermodynamic arguments to demonstrate that pure ionic liquids are expected to behave as dilute weak electrolyte solutions, with typical effective dissociated ion concentrations of less than 0.1% at room temperature. We performed equilibrium force–distance measurements across the common ionic liquid 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide ([C4mim][NTf2]) using a surface forces apparatus with in situ electrochemical control and quantitatively modeled these measurements using the van der Waals and electrostatic double-layer forces of the Derjaguin–Landau–Verwey–Overbeek theory with an additive repulsive steric (entropic) ion–surface binding force. Our results indicate that ionic liquids screen charged surfaces through the formation of both bound (Stern) and diffuse electric double layers, where the diffuse double layer is comprised of effectively dissociated ionic liquid ions. Additionally, we used the energetics of thermally dissociating ions in a dielectric medium to quantitatively predict the equilibrium for the effective dissociation reaction of [C4mim][NTf2] ions, in excellent agreement with the measured Debye length. Our results clearly demonstrate that, outside of the bound double layer, most of the ions in [C4mim][NTf2] are not effectively dissociated and thus do not contribute to electrostatic screening. We also provide a general, molecular-scale framework for designing ionic liquids with significantly increased dissociated charge densities via judiciously balancing ion pair interactions with bulk dielectric properties. Our results clear up several inconsistencies that have hampered scientific progress in this important area and guide the rational design of unique, high–free-ion density ionic liquids and ionic liquid blends. PMID:23716690

  2. 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

  3. Use of liquid membranes for treatment of nuclear wastes

    International Nuclear Information System (INIS)

    Dozol, J.F.

    1988-01-01

    The reprocessing operations produce liquid wastes in which the main components are nitric acid and sodium nitrate. The goal of the experiments is to separate trace amounts of radioactive elements from these acidic and high sodium nitrate content solutions. CMPO, a neutral bifunctional organophosphorus compound, and crown compounds (DC18 C6 - B21 C7) are able to extract respectively actinides, strontium and cesium from these high salinity solutions. The supported liquid membrane (SLM) render the use of expensive tailor-made extractant molecules like CMPO or crown ethers possible. The results obtained for the extraction of actinides and strontium are promising, but research must now be oriented towards improving the stability of the membrane

  4. Using benchmarking to minimize common DOE waste streams. Volume 1, Methodology and liquid photographic waste

    Energy Technology Data Exchange (ETDEWEB)

    Levin, V.

    1994-04-01

    Finding innovative ways to reduce waste streams generated at Department of Energy (DOE) sites by 50% by the year 2000 is a challenge for DOE`s waste minimization efforts. This report examines the usefulness of benchmarking as a waste minimization tool, specifically regarding common waste streams at DOE sites. A team of process experts from a variety of sites, a project leader, and benchmarking consultants completed the project with management support provided by the Waste Minimization Division EM-352. Using a 12-step benchmarking process, the team examined current waste minimization processes for liquid photographic waste used at their sites and used telephone and written questionnaires to find ``best-in-class`` industrv partners willing to share information about their best waste minimization techniques and technologies through a site visit. Eastman Kodak Co., and Johnson Space Center/National Aeronautics and Space Administration (NASA) agreed to be partners. The site visits yielded strategies for source reduction, recycle/recovery of components, regeneration/reuse of solutions, and treatment of residuals, as well as best management practices. An additional benefit of the work was the opportunity for DOE process experts to network and exchange ideas with their peers at similar sites.

  5. Liquid centrifugation for nuclear waste partitioning

    International Nuclear Information System (INIS)

    Bowman, C.D.

    1992-01-01

    The performance of liquid centrifugation for nuclear waste partitioning is examined for the Accelerator Transmutation of Waste Program currently under study at the Los Alamos National Laboratory. Centrifugation might have application for the separation of the LiF-BeF 2 salt from heavier radioactive materials fission product and actinides in the separation of fission product from actinides, in the isotope separation of fission-product cesium before transmutation of the 137 Cs and 135 Cs, and in the removal of spallation product from the liquid lead target. It is found that useful chemical separations should be possible using existing materials for the centrifuge construction for all four cases with the actinide fraction in fission product perhaps as low as 1 part in 10 7 and the fraction of 137 CS in 133 Cs being as low as a few parts in 10 5 . A centrifuge cascade has the advantage that it can be assembled and operated as a completely closed system without a waste stream except that associated with maintenance or replacement of centrifuge components

  6. Solidification of radioactive waste solutions by pelletization technique

    International Nuclear Information System (INIS)

    Akbar, A.H.; Koester, R.; Rudolph, G.

    1980-04-01

    A possible way of performing the cement fixation of radioactive wastes is the incorporation into cement pellets on a pan pelletizer, followed by embedding the pellets into an inactive cement matrix. This procedure is suitable for various types of waste, particularly for medium level liquid wastes, and can be used both at drum disposal and at in-situ solidification. This report describes some initial studies on the pelletization technique using a laboratory pelletizer. Formation and size of the pellets have been found to be determined by speed, angle, and load of the pan, ratio and mode of addition of the liquid and solid components, ect. Pellets in various compositions have been produced from cement and water or simulated waste solution, in some cases with the addition of bentonite for improving cesium retention. Some mechanical properties of the pellets such as fall height of fresh pellets, development of hardness (crush test), impact and abrasion resistance, have been determined. Some preliminary experiments were done on backfilling the void space between the pellets - about 40 per cent of the bulk volume - with cement grouts of appropriate compositions. (orig.) [de

  7. 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.)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  9. Microbial accumulation of uranium from nuclear liquid waste

    International Nuclear Information System (INIS)

    Mahmood, A.H.

    1986-01-01

    This investigation includes the isolation, identification and the fluctuations of the population densities of microorganisms in the nuclear liquid waste released by some laboratories of Iraqi Atomic Energy Commission. The efficiency of uranium accumulation on isolates (22 bacterial strains, 24 fungal strains and 6 yeast strains) was assessed in aqueous solution using fluorometric techniques. Two of the isolated microoganisms namely Bacillus sp. -15B and Mucor sp.16F showed exceptionally high attitude towards uranium accumulation. Optimal conditions required for efficient accumulation and recovery of uranium was then studied using the two selected isolates. 10 figs.; 162 refs.; 16 tabs

  10. Use of ferric- and ferrous-salts in liquid waste treatment processes

    International Nuclear Information System (INIS)

    Efremenkov, V.M.; Toropov, I.G.; Toropova, V.V.; Satsukevich, V.M.; Davidov, J.P.; Jabrodsky, V.N.; Prokshin, N.E.

    1995-01-01

    Treatment of spent decontamination solutions is the most complicated task in the whole problem of management of liquid radioactive waste, because quite often they have complex compositions, which makes it difficult to find for them effective and non-expensive treatment technology. New methods of treatment of such a waste is proposed based on use of specific sorption ability of ferro- and ferri-species in solution. These species are often present in solution as the by-products, and in combination with other components of decontamination solution they can be used as initial substances for synthesis of valuable sorbents directly in treating solution. Using specific compositions and conditions in solution, it is possible to make liquid waste treatment process more effective and less expensive. Particular examples of this process is presented in this work

  11. Method of solidifying liquid radioactive wastes

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. Device for concentrating radioactive liquid wastes

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  13. 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)

  14. 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.

  15. 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

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

    International Nuclear Information System (INIS)

    Kumar, Surender; Jain, Savita; Raj, Kanwar

    2010-01-01

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

  17. Ionic liquids, electrolyte solutions including the ionic liquids, and energy storage devices including the ionic liquids

    Science.gov (United States)

    Gering, Kevin L.; Harrup, Mason K.; Rollins, Harry W.

    2015-12-08

    An ionic liquid including a phosphazene compound that has a plurality of phosphorus-nitrogen units and at least one pendant group bonded to each phosphorus atom of the plurality of phosphorus-nitrogen units. One pendant group of the at least one pendant group comprises a positively charged pendant group. Additional embodiments of ionic liquids are disclosed, as are electrolyte solutions and energy storage devices including the embodiments of the ionic liquid.

  18. Processing of waste solutions from electrochemical decontamination

    International Nuclear Information System (INIS)

    Charlot, L.A.; Allen, R.P.; Arrowsmith, H.W.; Hooper, J.L.

    1979-09-01

    The use of electropolishing as a decontamination technique will be effective only if we can minimize the amount of secondary waste requiring disposal and economically recycle part of the decontamination electrolyte. Consequently, a solution purification method is needed to remove the dissolved contamination and metal in the electrolyte. This report describes the selection of a purification method for a phosphoric acid electrolyte from the following possible acid reclamation processes: ion exchange, solvent extraction, precipitation, distillation, electrolysis, and membrane separation

  19. Treatment for hydrazine-containing waste water solution

    Science.gov (United States)

    Yade, N.

    1986-01-01

    The treatment for waste solutions containing hydrazine is presented. The invention attempts oxidation and decomposition of hydrazine in waste water in a simple and effective processing. The method adds activated charcoal to waste solutions containing hydrazine while maintaining a pH value higher than 8, and adding iron salts if necessary. Then, the solution is aerated.

  20. The waste management program VUB-AZ: An integrated solution for nuclear biomedical waste management

    International Nuclear Information System (INIS)

    Covens, P.; Sonck, M.; Eggermont, G.; Meert, D.

    2001-01-01

    Due to escalating costs and the lack of acceptance of near-surface disposal facilities, the University of Brussels (VUB) and its Academic hospital (AZ) have developed an on-site waste storage program in collaboration with Canberra Europe. This programme is based on selective collection, measurement before decay, storage for decay of short-lived radionuclides, measurement after decay and eventual clearance as non-nuclear waste. It has proved its effectiveness over the past 5 years. Effective characterisation for on-site storage for decay of short-lived radionuclides makes selective collection of waste streams mandatory and requires motivated and trained laboratory staff. Dynamic optimisation of this selective collection increases the efficiency of the storage for decay programme. The accurate qualitative and quantitative measurement of nuclear biomedical waste before decay has several advantages such as verification of correct selective collection, optimisation of the decay period and possibility of clearance below the minimal detectable activity. In the research phase of the program several measurement techniques were investigated. The following measurement concept was selected. Closed PE drums containing low density solid waste materials contaminated with small amounts of β/γ-or pure β-emitting radionuclides are assessed for specific activity by the Canberra measurement unit for nuclear biomedical waste, based on a HPGe-detector. Liquid waste containing (β/γ-emitters are characterised by the same technique while for pure β-emitting liquid waste a Packard liquid scintillation counter is used. Measurement results are obtained by using the gamma-spectroscopy software Genie-2000. A user-friendly interface, based on Procount-2000 and optimised by Canberra for the characterisation of nuclear biomedical waste, has increased the sample throughput of the measurement concept. The MDA (minimal detectable activity) of different radionuclides obtained by the measurement

  1. Future radioactive liquid waste streams study

    International Nuclear Information System (INIS)

    Rey, A.S.

    1993-11-01

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

  2. Future radioactive liquid waste streams study

    Energy Technology Data Exchange (ETDEWEB)

    Rey, A.S.

    1993-11-01

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

  3. Aerogels from Chitosan Solutions in Ionic Liquids

    Directory of Open Access Journals (Sweden)

    Gonzalo Santos-López

    2017-12-01

    Full Text Available Chitosan aerogels conjugates the characteristics of nanostructured porous materials, i.e., extended specific surface area and nano scale porosity, with the remarkable functional properties of chitosan. Aerogels were obtained from solutions of chitosan in ionic liquids (ILs, 1-butyl-3-methylimidazolium acetate (BMIMAc, and 1-ethyl-3-methyl-imidazolium acetate (EMIMAc, in order to observe the effect of the solvent in the structural characteristics of this type of materials. The process of elaboration of aerogels comprised the formation of physical gels through anti-solvent vapor diffusion, liquid phase exchange, and supercritical CO2 drying. The aerogels maintained the chemical identity of chitosan according to Fourier transform infrared spectrophotometer (FT-IR spectroscopy, indicating the presence of their characteristic functional groups. The internal structure of the obtained aerogels appears as porous aggregated networks in microscopy images. The obtained materials have specific surface areas over 350 m2/g and can be considered mesoporous. According to swelling experiments, the chitosan aerogels could absorb between three and six times their weight of water. However, the swelling and diffusion coefficient decreased at higher temperatures. The structural characteristics of chitosan aerogels that are obtained from ionic liquids are distinctive and could be related to solvation dynamic at the initial state.

  4. Low-level liquid waste decontamination by inorganic ion exchange

    International Nuclear Information System (INIS)

    Campbell, D.O.; Lee, D.D.; Dillow, T.A.

    1990-01-01

    Improved processes are being developed to treat contaminated liquid wastes that have been and continue to be generated at Oak Ridge National Laboratory. The most serious contaminants are 137 Cs and 90 Sr, and certain inorganic ion-exchange material have given promising results. Nickel and cobalt hexacyanoferrate (II) compounds are extremely selective for cesium removal, with distribution coefficients in excess of 10 6 even in the presence of high cesium and moderate potassium concentrations. Sodium titanate is selective for strontium removal from solutions with high alkali metal concentrations, especially at high pH. These separations are so efficient that one or two stages of simple, batch separation can yield large DFs (∼10 4 ) while still generating small volumes of solid waste

  5. 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)

  6. Vitrification of high-level liquid wastes

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Treatment of fast reactor liquid waste- electrochemical method

    International Nuclear Information System (INIS)

    Mahato, Swapan Kumar; Sudha, R.; Anthonysamy, S.; Muralidaran, P.

    2015-01-01

    During the operation of fast reactors, components get wetted by sodium. The sodium wetted primary components such as pumps and intermediate heat exchangers (IHX) in fast reactors are cleaned free of sodium followed by suitable chemical decontamination process before taking them for maintenance or for disposal. This helps in reduction of radiation dose to the operating personnel. Sodium cleaning and decontamination generates large volumes of liquid effluent. The activity in the liquid effluent during sodium cleaning/decontamination is due to 22 Na, 54 Mn, 58 Co, 60 Co, 59 Fe, 137 Cs and 134 Cs. It is required to chemically treat the effluent to reduce the activity levels prior to storage in tanks and transportation to the waste management facility for final disposal. Conventionally the ion exchange method is used for removal of radionuclides which produces large quantities of secondary waste. A method which is suitable both for removal of radionuclides present in low concentration and that avoids generation of large quantities of secondary waste is required. Hence an electrochemical method for metal ion removal is attempted in this work which produces little or no secondary waste. Electrochemical method towards removal of manganese ions was finalized earlier using reticulated vitreous carbon (RVC) from simulated decontamination solution containing a mixture of sulphuric and phosphoric acids. In continuation of the experiments for the removal of cesium ions from simulated cleaning solution which has an alkaline pH, a thin film of nickel hexacyanoferrate (NiHCF) was deposited electrochemically on the surface of RVC. Hexacyanoferrates are known for selectively binding cesium. This NiHCF coated RVC was used for electrodeposition of Cs ions. NiHCF coated and Cs deposited RVC was characterized using SEM/EDX analysis. EDX analysis confirms the presence of Cs on NiHCF coated RVC. (author)

  8. 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

  9. The Sonophysics and Sonochemistry of Liquid Waste Quantification and Remediation

    Energy Technology Data Exchange (ETDEWEB)

    Matula, Thomas J.

    1998-06-01

    This research is being conducted to (a) perform an in-depth and comprehensive study of the fundamentals of acoustic cavitation and nonlinear bubble dynamics, (b) elucidate the fundamental physics of sonochemical reactions, (c) examine the potential of sonoluminescence to quantify and monitor the presence of alkali metals and other elements in waste liquids, (d) design and evaluate more effective sonochemical reactors for waste remediation, and (e) determine the optimal acoustical parameters in the use of sonochemistry for liquid-waste-contaminant remediation. So far cells have been designed for multibubble sonoluminescence (MBSL) and single-bubble sonoluminescence (SBSL) spectroscopy experiments. Positive results have been obtained in both systems using a Raman system which covers the wavelength range from 790 to 1,070 nm. Further progress from year-1 involved the use of the newly discovered technique of changing the pressure head above the cavitation field to increase the light emission from MBSL. A second method for changing the pressure head involves pressure-jumping, whereby the pressure in the head space above the solution is quickly increased to a new steady value.

  10. Recovering low-turbidity cutting liquid from silicon slurry waste.

    Science.gov (United States)

    Tsai, Tzu-Hsuan; Shih, Yu-Pei

    2014-04-30

    In order to recover a low-turbidity polyalkylene glycol (PAG) liquid from silicon slurry waste by sedimentation, temperatures were adjusted, and acetone, ethanol or water was used as a diluent. The experimental results show that the particles in the waste would aggregate and settle readily by using water as a diluent. This is because particle surfaces had lower surface potential value and weaker steric stabilization in PAG-water than in PAG-ethanol or PAG-acetone solutions. Therefore, water is the suggested diluent for recovering a low-turbidity PAG (sedimentation. After 50 wt.% water-assisted sedimentation for 21 days, the solid content of the upper liquid reduced to 0.122 g/L, and the turbidity decreased to 44 NTU. The obtained upper liquid was then vacuum-distillated to remove water. The final recovered PAG with 0.37 NTU had similar viscosity and density to the unused PAG and could be reused in the cutting process. Copyright © 2014 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. 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.)

  13. Membrane technologies for liquid radioactive waste treatment

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Decontamination of waste radioactive polluted solutions in radiation treatment

    International Nuclear Information System (INIS)

    Simova, G.; Boyadzhiev, A.; Mikhajlov, M.G.; Shopov, N.

    1979-01-01

    The decontamination capacity of solutions of the trivial cleaning Bulgarian preparations ''Mipro'', ''Sana'', ''Synthek'' and ''Univer'' for different surfaces (steel, glass, PVC and linoleum) contaminated with cesium-134, strontium-85 or cerium-144 chlorides, was studied. Concentrations from 5 to 15 g/l of the solutions used in this study displayed a degree of cleaning over 90%. Higher concentration of the solution does not improve its cleaning capacity. For evaluation of foam formation by the solutions, the so called ''foam column stability coefficient'' has been adopted. This coefficient represents the ratio between the height of the foam column and the time of its half life, referred to the time for the foam column formation when blown through with a constant air current. On the basis of this index, solutions of the preparation ''Mipro'' proved to be the best ones for decontamination - in the whole investigated concentration span, the foam column stability coefficient for the solutions of this preparation is with two orders lower than the respective coefficient of the other preparations. It was experimentally established that radiation treatment of radio-contaminated solutions reduces the foam column stability coefficient. Radiation treatment should be carried out in a gamma field, realizing at least one megarad within an acceptable for the liquid wastes time period. (A.B.)

  15. TECHNICAL NOTE LIQUID WASTE DISPOSAL IN URBAN LOW ...

    African Journals Online (AJOL)

    In the ideal case the liquid waste can safely be disposed of in a properly designed and integrated network of pipes, which collect and transmit the liquid waste into a treatment plant. However, such a system is costly and needs a substantial amount of initial investment to start operating and subsequently to maintain.

  16. Effect of municipal liquid waste on corrosion susceptibility of ...

    African Journals Online (AJOL)

    This investigation studied the effect of municipal liquid waste discharged into the environment within Kano municipal area on the corrosion susceptibility of galvanized steel pipe burial underground. Six stagnant and six moving municipal liquid waste samples were used for the investigation. The corrosion rate of the ...

  17. Liquid waste treatment at plutonium fuels fabrication facility, 2

    International Nuclear Information System (INIS)

    Matsumoto, Ken-ichi; Itoh, Ichiroh; Ohuchi, Jin; Miyo, Hiroaki

    1974-01-01

    The economics in the management of the radioactive liquid waste from Plutonium Fuels Fabrication Facility with sludge-blanket type flocculators has been evaluated. (1) Cost calculation: The cost of chemicals and electricity to treat 1 cubic meter of liquid waste is about 876 yen, while the total operating cost is 250 thousand yen per cubic meter in the case of 140 m 3 /year treatment. These figures are much higher than those for ordinary wastes, due to the particular operation against plutonium. (2) Proposal of the closed system for liquid waste treatment at PFFF: In the case of a closed system using evaporator, ion exchange column and rotary-kiln calciner, the operating cost is estimated at 40 thousand yen per cubic meter of liquid waste. Final radioactivity of treated liquid is below 10 -8 micro curies/ml. (Mori, K.)

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

    International Nuclear Information System (INIS)

    Kwak, Kyung Kil; Ji, Young Yong

    2010-12-01

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

  19. 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)

  20. A process for solidifying radioactive liquid waste

    International Nuclear Information System (INIS)

    Mergan, L.M.; Cordier, J.-P.

    1981-01-01

    In a process for solidifying radioactive liquid waste, its pH is adjusted, solids precipitated and then it is concentrated to about 50% solids content using a thin film evaporator, the concentrate then being dried to powder in a heated mixer. The mixer has a heated wall and working means, e.g. a rotor and helical screw, to shear the dried concentrate from the internal walls, subdivide it into a dry particulate powder, and advance the powder to the mixer outlet. The dried particles are then encapsulated in a suitable matrix. Vapour from the mixer and evaporator is condensed and recycled after any particles have been removed from it. The mixer may both dry the concentrate and mix the dry particles with the encapsulating matrix, and possibly, part of the mixer may be used for pH adjustment and precipitation. (author)

  1. 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.)

  2. Cement encapsulation of low-level waste liquids. Final report

    International Nuclear Information System (INIS)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place

  3. UKAEA contract no. 3: miscellaneous solid, liquid and gaseous wastes

    International Nuclear Information System (INIS)

    Partridge, B.A.

    1984-12-01

    This document reports work carried out in 1982/83 on the following topics concerned with the treatment and disposal of intermediate level wastes: flowsheeting; dewatering low and medium level radioactive wastes; applications of ultrafiltration in the treatment of radioactive liquid wastes; ion exchange processes; electrical processes for the treatment of medium active liquid wastes; chemical conversion of Zircaloy cladding to oxide; fast reactor fuel element cladding; dissolver residues; fuel cladding and ion exchanger immobilisation - radioactive trials; thermal techniques; development and assessment of medium level waste forms. (U.K.)

  4. Liquid waste management: The case of Bahir Dar, Ethiopia ...

    African Journals Online (AJOL)

    Background: Human beings pollute the environment with their industrial and domestic wastes. In Bahir Dar Town there is no conventional municipal waste water collection and treatment system. Objective: The aim of this study was to describe the liquid waste disposal practices of the residents of Bahir Dar Town and to ...

  5. Selection of Technical Solutions for the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    2017-07-01

    The objectives of this publication are to identify and critically review the criteria to be considered while selecting waste management technologies; summarize, evaluate, rank and compare the different technical solutions; and offer a systematic approach for selecting the best matching solution. This publication covers the management of radioactive waste from all nuclear operations, including waste generated from research reactors, power reactors, and nuclear fuel cycle activities including high level waste (HLW) arising from reprocessing and spent nuclear fuel declared as waste (SFW), as well as low level waste (LLW) and intermediate level waste (ILW) arising from the production and use of radionuclides in industry, agriculture, medicine, education and research.

  6. 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

  7. Liquid level measurement in high level nuclear waste slurries

    International Nuclear Information System (INIS)

    Weeks, G.E.; Heckendorn, F.M.; Postles, R.L.

    1990-01-01

    Accurate liquid level measurement has been a difficult problem to solve for the Defense Waste Processing Facility (DWPF). The nuclear waste sludge tends to plug or degrade most commercially available liquid-level measurement sensors. A liquid-level measurement system that meets demanding accuracy requirements for the DWPF has been developed. The system uses a pneumatic 1:1 pressure repeater as a sensor and a computerized error correction system. 2 figs

  8. 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)

  9. Liquid waste disposal and reuse of waste water; Smaltimento e riuso delle acque reflue

    Energy Technology Data Exchange (ETDEWEB)

    Indelicato, S. [Catania Univ. (Italy). Cattedra di Idraulica Agraria; De Dominicis, G. [S.M.T. Societa Mineraria Trasimeno s.p.a.- Gruppo ACEA, Rome (Italy)

    1996-03-01

    The disposal of liquid wastes determine an environmental impact. Waste processing plants reduce this impact but, in case of malfunction or scheduled maintenance are emitted aerosols, odors and noise. Mitigation of this effects is possible with coverage or plants screen.

  10. Concepts for detritiation of waste liquids

    International Nuclear Information System (INIS)

    King, C.M.; Van Brunt, V.; Garber, A.R.; King, R.B.

    1991-01-01

    Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the SRS reactors, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, so research and development programs have been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R ampersand D efforts world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in future fusion reactors. This paper presents a review of some of the methods that have been proposed, studied, and developed for removal of tritium from light and heavy water, along with some new concepts for aqueous detritiation directly from liquid oxide (HTO) bearing feed streams

  11. Liquid waste handling facilities for a conceptual LWR spent fuel reprocessing complex

    International Nuclear Information System (INIS)

    Witt, D.C.; Bradley, R.F.

    1978-01-01

    The waste evaporator systems and the methods for evaporating the liquid wastes of various radioactivity levels are discussed. After the liquid wastes are evaporated and nitric acid is recovered the high-level liquid waste is incorporated into borosilicate glass and the intermediate-level liquid waste into concrete for final disposal

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. Newly Generated Liquid Waste Processing Alternatives Study, Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Landman, William Henry; Bates, Steven Odum; Bonnema, Bruce Edward; Palmer, Stanley Leland; Podgorney, Anna Kristine; Walsh, Stephanie

    2002-09-01

    This report identifies and evaluates three options for treating newly generated liquid waste at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory. The three options are: (a) treat the waste using processing facilities designed for treating sodium-bearing waste, (b) treat the waste using subcontractor-supplied mobile systems, or (c) treat the waste using a special facility designed and constructed for that purpose. In studying these options, engineers concluded that the best approach is to store the newly generated liquid waste until a sodium-bearing waste treatment facility is available and then to co-process the stored inventory of the newly generated waste with the sodium-bearing waste. After the sodium-bearing waste facility completes its mission, two paths are available. The newly generated liquid waste could be treated using the subcontractor-supplied system or the sodium-bearing waste facility or a portion of it. The final decision depends on the design of the sodium-bearing waste treatment facility, which will be completed in coming years.

  15. Addition of liquid waste incineration capability to the INEL's low-level waste incinerator

    International Nuclear Information System (INIS)

    Steverson, E.M.; Clark, D.P.; McFee, J.N.

    1986-01-01

    A liquid waste system has recently been installed in the Waste Experimental Reduction Facility (WERF) incinerator at the Idaho National Engineering Laboratory (INEL). In this paper, aspects of the incineration system such as the components, operations, capabilities, capital cost, EPA permit requirements, and future plans are discussed. The principal objective of the liquid incineration system is to provide the capability to process hazardous, radioactively contaminated, non-halogenated liquid wastes. The system consists primarily of a waste feed system, instrumentation and controls, and a liquid burner, which were procured at a capital cost of $115,000

  16. 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

  17. Treatment of ORNL liquid low-level waste

    International Nuclear Information System (INIS)

    Berry, J.B.; Brown, C.H. Jr.; Fowler, V.L.; Robinson, S.M.

    1988-01-01

    Discontinuation of the hydrofracture disposal method at Oak Ridge National Laboratory (ORNL) has caused intensive efforts to reduce liquid waste generation. Improving the treatment of slightly radioactive liquid waste, called process waste, has reduced the volume of the resulting contaminated liquid radioactive waste effluent by 66%. Proposed processing improvements could eliminate the contaminated liquid effluent and reduce solid low-level waste by an additional one-third. The improved process meets stringent discharge limits for radionuclides. Discharge limits for radionuclides are expected to be enforced at the outfall of the treatment plant to a creek; currently, limits are enforced at the reservation boundary. Plant discharge is monitored according to the National Pollutant Discharge Elimination System (NPDES) permit for ORNL. 1 ref., 4 figs., 2 tabs

  18. 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

  19. removal of hazardous pollutants from industrial waste solutions using membrane techniques

    International Nuclear Information System (INIS)

    Selim, Y.T.M.

    2001-01-01

    the removal of hazardous pollutants from industrial waste solutions is of essential demand field for both scientific and industrial work. the present work includes detailed studies on the possible use of membrane technology especially liquid emulsion membrane for the removal of hazardous pollutants such as; cadmium , cobalt , lead, copper and uranium from different industrial waste solution . this research can be applied for mixed waste problems. the work carried out in this thesis is presented in three main chapters, namely introduction, experimental and results and discussion

  20. Decontamination of liquid radioactive waste by thorium phosphate

    International Nuclear Information System (INIS)

    Rousselle, J.; Grandjean, S.; Dacheux, N.; Genet, M.

    2004-01-01

    In the field of the complete reexamination of the chemistry of thorium phosphate and of the improvement of the homogeneity of Thorium Phosphate Diphosphate (TPD, Th 4 (PO 4 ) 4 P 2 O 7 ) prepared at high temperature, several crystallized compounds were prepared as initial powdered precursors. Due to the very low solubility products associated to these phases, their use in the field of the efficient decontamination of high-level radioactive liquid waste containing actinides (An) was carefully considered. Two main processes (called 'oxalate' and 'hydrothermal' chemical routes) were developed through a new concept combining the decontamination of liquid waste and the immobilization of the actinides in a ceramic matrix (TPD). In phosphoric media ('hydrothermal route'), the key-precursor was the Thorium Phosphate Hydrogen Phosphate hydrate (Th 2 (PO 4 ) 2 (HPO 4 ). H 2 O, TPHP, solubility product log(K S,0 0 ) ∼ - 67). The replacement of thorium by other tetravalent actinides (U, Np, Pu) in the structure, leading to the preparation of Th 2-x/2 An x/2 (PO 4 ) 2 (HPO 4 ). H 2 O solid solutions, was examined. A second method was also considered in parallel to illustrate this concept using the more well-known precipitation of oxalate as the initial decontamination step. For this method, the final transformation to single phase TPD containing actinides was purchased by heating a mixture of phosphate ions with the oxalate precipitate at high temperature. (authors)

  1. Liquid radioactive waste processing improvement of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes

    2005-01-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  2. The TRUEX [TRansUranium EXtraction] process and the management of liquid TRU [transuranic] waste

    International Nuclear Information System (INIS)

    Schulz, W.W.; Horwitz, E.P.

    1987-01-01

    The TRUEX process is a new generic liquid-liquid extraction process for removal of all actinides from acidic nitrate or chloride nuclear waste solutions. Because of its high efficiency and great flexibility, the TRUEX process appears destined to be widely used in the US and possibly in other countries for cost-effective management and disposal of transuranic (TRU) wastes. In the US, TRU wastes are those that contain ≥3.7 x 10 6 Bq/kg) of TRU elements with half-lives greater than 20 y. This paper gives a brief review of the relevant chemistry and summarizes the current status of development and deployment of the TRUEX (TRansUranium EXtraction) process flowsheets to treat specific acidic waste solutions at several US Department of Energy sites. 19 refs., 4 figs., 4 tabs

  3. Solidification of low-level radioactive liquid waste using a cement-silicate process

    International Nuclear Information System (INIS)

    Grandlund, R.W.; Hayes, J.F.

    1979-01-01

    Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

  4. Nuclear waste management. Pioneering solutions from Finland

    International Nuclear Information System (INIS)

    Rasilainen, Kari

    2016-01-01

    Presentation outline: Background: Nuclear energy in Finland; Nuclear Waste Management (NWM) Experiences; Low and Intermediate Level Waste (LILW); High Level Waste - Deep Geological Repository (DGR); NWM cost estimate in Finland; Conclusions: World-leading expert services

  5. 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.

  6. 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.

  7. Concentration of High Level Radioactive Liquid Waste. Basic data acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Juvenelle, A.; Masson, M.; Garrido, M.H. [DEN/VRH/DRCP/SCPS/LPCP, BP 17171 - 30207 Bagnols sur Ceze Cedex (France)

    2008-07-01

    Full text of publication follows: In order to enhance its knowledge about the concentration of high level liquid waste (HLLW) from the nuclear fuel reprocessing process, a program of studies was defined by Cea. In a large field of acidity, it proposes to characterize the concentrated solution and the obtained precipitates versus the concentration factor. Four steps are considered: quantification of the salting-out effect on the concentrate acidity, acquisition of solubility data, precipitates characterisation versus the concentration factor through aging tests and concentration experimentation starting from simulated fission products solutions. The first results, reported here, connect the acidity of the concentrated solution to the concentration factor and allow us to precise the field of acidity (4 to 12 N) for the next experiments. In this field, solubility data of various elements (Ba, Sr, Zr...) are separately measured at room temperature, in nitric acid in a first time, then in the presence of various species present in medium (TBP, PO{sub 4}{sup 3-}). The reactions between these various elements are then investigated (formation of insoluble mixed compounds) by following the concentration cations in solution and characterising the precipitates. (authors)

  8. Electrolytic treatment of liquid waste containing ammonium nitrate

    International Nuclear Information System (INIS)

    Komori, R.; Ogawa, N.; Ohtsuka, K.; Ohuchi, J.

    1981-01-01

    A study was made on the safe decomposition of ammonium nitrate, which is the main component of α-liquid waste from plutonium fuel facilities, by means of electrolytic reduction and thermal decomposition. In the first stage, ammonium nitrate is reduced to ammonium nitrite by electrolytic reduction using an electrolyser with a cation exchange membrane as a diaphragm. In the second stage, ammonium nitrite is decomposed to N 2 and H 2 O. The alkaline region and a low temperature are preferable for electrolytic reduction and the acidic region and high temperature for thermal decomposition. A basis was established for an ammonium nitrate treatment system in aqueous solution through the operation of a bench-scale unit, and the operating data obtained was applied to the basic design of a 10-m 3 /a facility. (author)

  9. Design and operation of off-gas cleaning systems at high level liquid waste conditioning facilities

    International Nuclear Information System (INIS)

    1988-01-01

    The immobilization of high level liquid wastes from the reprocessing of irradiated nuclear fuels is of great interest and serious efforts are being undertaken to find a satisfactory technical solution. Volatilization of fission product elements during immobilization poses the potential for the release of radioactive substances to the environment and necessitates effective off-gas cleaning systems. This report describes typical off-gas cleaning systems used in the most advanced high level liquid waste immobilization plants and considers most of the equipment and components which can be used for the efficient retention of the aerosols and volatile contaminants. In the case of a nuclear facility consisting of several different facilities, release limits are generally prescribed for the nuclear facility as a whole. Since high level liquid waste conditioning (calcination, vitrification, etc.) facilities are usually located at fuel reprocessing sites (where the majority of the high level liquid wastes originates), the off-gas cleaning system should be designed so that the airborne radioactivity discharge of the whole site, including the emission of the waste conditioning facility, can be kept below the permitted limits. This report deals with the sources and composition of different kinds of high level liquid wastes and describes briefly the main high level liquid waste solidification processes examining the sources and characteristics of the off-gas contaminants to be retained by the off-gas cleaning system. The equipment and components of typical off-gas systems used in the most advanced (large pilot or industrial scale) high level liquid waste solidification plants are described. Safety considerations for the design and safe operation of the off-gas systems are discussed. 60 refs, 31 figs, 17 tabs

  10. Corrosion of a carbon steel in simulated liquid nuclear wastes

    International Nuclear Information System (INIS)

    Saenz Gonzalez, Eduardo

    2005-01-01

    This work is part of a collaboration agreement between CNEA (National Atomic Energy Commission of Argentina) and USDOE (Department of Energy of the United States of America), entitled 'Tank Corrosion Chemistry Cooperation', to study the corrosion behavior of carbon steel A537 class 1 in different simulated non-radioactive wastes in order to establish the safety concentration limits of the tank waste chemistry at Hanford site (Richland-US). Liquid high level nuclear wastes are stored in tanks made of carbon steel A537 (ASTM nomenclature) that were designed for a service life of 20 to 50 years. A thickness reduction of some tank walls, due to corrosion processes, was detected at Hanford site, beyond the existing predicted values. Two year long-term immersion tests were started using non radioactive simulated liquid nuclear waste solutions at 40 C degrees. This work extends throughout the first year of immersion. The simulated solutions consist basically in combinations of the 10 most corrosion significant chemical components: 5 main components (NaNO 3 , NaCl, NaF, NaNO 2 and NaOH) at three concentration levels and 5 secondary components at two concentration levels. Measurements of the general corrosion rate with time were performed for carbon steel coupons, both immersed in the solutions and in the vapor phases, using weight loss and electrochemistry impedance spectroscopy techniques. Optic and scanning electron microscopy examination, analysis of U-bend samples and corrosion potential measurements, were also done. Localized corrosion susceptibility (pitting and crevice corrosion) was assessed in isolated short-term tests by means of cyclic potentiodynamic polarization curves. The effect of the simulated waste composition on the corrosion behavior of A537 steel was studied based on statistical analyses. The Surface Response Model could be successfully applied to the statistical analysis of the A537 steel corrosion in the studied solutions. General corrosion was not

  11. Oxidizing purification of liquid radioactive waste from organic substances and radionuclides by K permanganate

    International Nuclear Information System (INIS)

    Rudenko, L.I.; Dzhuzha, O.V.; Khan, V.E.

    2007-01-01

    The basic opportunity of the oxidizing purification of liquid radioactive waste (LRW) with the use of a water solution of potassium permanganate for the preliminary preparation of LRW at a stage prior to the evaporating devices of the Chernobyl NPP is shown

  12. Liquid return from gas pressurization of grouted waste

    International Nuclear Information System (INIS)

    Powell, W.J.; Benny, H.L.

    1994-05-01

    The ability to force pore liquids out of a simulated waste grout matrix using air pressure was measured. Specimens cured under various conditions were placed in a permeameter and subjected to increasing air pressure. The pressure was held constant for 24 hours and then stepped up until either liquid was released or 150 psi was reached. One specimen was taken to 190 psi with no liquid release. Permeability to simulated tank waste was then measured. Compressive strength was measured following these tests. This data is to assess the amount of fluid that might be released from grouted waste resulting from the buildup of radiolytically generated hydrogen and other gasses within the waste form matrix. A plot of the unconfined compressive strength versus breakthrough pressures identifies a region of ''good'' grout, which will resist liquid release

  13. 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.)

  14. 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.)

  15. Synergistic extraction behaviour of americium from simulated acidic waste solutions

    International Nuclear Information System (INIS)

    Pathak, P.N.; Veeraraghavan, R.; Mohapatra, P.K.; Manchanda, V.K.

    1998-01-01

    The extraction behaviour of americium has been investigated with mixtures of 3-phenyl-4-benzoyl-5-isoxazolone (PBI) and oxodonors viz. tri-n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO) and di-n-butyl octanamide (DBOA) using dodecane as the diluent from 1-6 M HNO 3 media. It is observed that D Am remains unaltered with PBI concentration (in the range 0.06-0.1 M) at 1.47 M TBP in the entire range of HNO 3 concentration. PBI and TBP in combination appears more promising compared to other synergistic systems. The possibility of using this mixture for americium removal from high level liquid waste solution has been explored. Extraction studies indicated that prior removal of uranium by 20% TBP in dodecane is helpful in the quantitative recovery of americium in three contacts. Effect of lanthanides on D Am is found to be marginal. (orig.)

  16. 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

  17. Evaluation of mercury in the liquid waste processing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vijay [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shah, Hasmukh [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Occhipinti, John E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, Richard E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-13

    This report provides a summary of Phase I activities conducted to support an Integrated Evaluation of Mercury in Liquid Waste System (LWS) Processing Facilities. Phase I activities included a review and assessment of the liquid waste inventory and chemical processing behavior of mercury using a system by system review methodology approach. Gaps in understanding mercury behavior as well as action items from the structured reviews are being tracked. 64% of the gaps and actions have been resolved.

  18. 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

  19. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  20. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  1. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. The different solutions for the waste storage

    International Nuclear Information System (INIS)

    Fillion, E.

    2001-01-01

    Created in 1979, the National agency for the management of radioactive waste (A.N.D.R.A.) is a public establishment in charge of the management of radioactive waste produced in France. It is independent from waste producers and watches over the long term protection of man and his environment, at any step of radioactive waste management. It has for mission to check the waste quality and to conceive, to establish, to build and to manage storage centers where waste are stored according their characteristics. (N.C.)

  3. Carbon Market and Integrated Waste Solutions : a Case Study of ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Carbon Market and Integrated Waste Solutions : a Case Study of Indonesia ... dual purpose of helping developing countries achieve sustainable development ... with a view to devising integrated waste management solutions in urban centres ... and disseminate them through national, regional and international networks.

  4. Treatment of organic waste solutions containing tributyl phosphate

    International Nuclear Information System (INIS)

    Drobnik, S.

    The two processes developed in the laboratory for treating waste solutions containing TBP, namely TBP separation with phosphoric acid and saponification were tested on a semi-industrial scale. A waste solution from the first phase of the Karlsruhe reprocessing plant was used

  5. APPLICATION OF PULSE COMBUSTION TO INCINERATION OF LIQUID HAZARDOUS WASTE

    Science.gov (United States)

    The report gives results of a study to determine the effect of acoustic pulsations on the steady-state operation of a pulse combustor burning liquid hazardous waste. A horizontal tunnel furnace was retrofitted with a liquid injection pulse combustor that burned No. 2 fuel oil. Th...

  6. Energy from waste: a wholly acceptable waste-management solution

    International Nuclear Information System (INIS)

    Porteous, A.

    1997-01-01

    This paper briefly reviews the 'waste management hierarchy' and why it should be treated as a checklist and not a piece of unquestioning dogma. The role of energy from waste (EfW) is examined in depth to show that it is a rigorous and environmentally sound waste-management option which complements other components of the waste-management hierarchy and assists resource conservation. (Copyright (c) 1997 Elsevier Science B.V., Amsterdam. All rights reserved.)

  7. Radioactive waste management - a safe solution

    International Nuclear Information System (INIS)

    1993-01-01

    This booklet sets out current United Kingdom government policy regarding radioactive waste management and is aimed at reassuring members of the public concerned about the safety of radioactive wastes. The various disposal or, processing or storage options for low, intermediate and high-level radioactive wastes are explained and sites described, and the work of the Nuclear Industry Radioactive Waste Executive (NIREX) is outlined. (UK)

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper documents the status of the technical design for the upgrade and expansion to the existing Low-level Liquid Radioactive Waste (LLLRW) treatment facility in Murmansk, the Russian Federation. This facility, owned by the Ministry of Transportation and operated by the Russian company RTP Atomflot in Murmansk, Russia, has been used by the Murmansk Shipping Company (MSCo) to process low-level liquid radioactive waste generated by the operation of its civilian icebreaker fleet. The purpose of the new design is to enable Russia to permanently cease the disposal at sea of LLLRW in the Arctic, and to treat liquid waste and high saline solutions from both the Civil and North Navy Fleet operations and decommissioning activities. Innovative treatments are to be used in the plant which are discussed in this paper

  12. Denitration and chemical precipitation of medium level liquid wastes and conditioning of high level wastes from low level liquid wastes by a roll dryer and subsequent vitrification

    International Nuclear Information System (INIS)

    Halaszovich, S.; Dix, S.; Harms, R.

    1987-01-01

    Medium level liquid waste (MAW) from the reprocessing need after being fixed in cement an additional shielding to meet required radiation limits for handling and transportation. Normally this shielding consists of concrete and its weight and volume is several times higher than that of the waste product itself. By means of caesium separation using nickel-potassium-hexacyanoferrate and after few years of interim storage waiting for the decay of Ruthenium and Antimony the activities will be reduced below permissible values. (13 MBq/l in waste solution for Cs, 28 MBq/l for Sb and 34 MBq/l for Ru). Below these limits there is no need for additional shielding after cementation in a 400 l drum. Experimental results show, that Caesium can be precipitated and separated effectively not only in laboratory but also in a larger scale under hot cell conditions. The process investigated in this work has been developed from the FIPS process for vitrification of highly radioactive fission product solutions. It consists of: denitration, precipitation, sludge separation, drying and melting

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

    International Nuclear Information System (INIS)

    Ludwieg, F.

    1976-01-01

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

  14. Operating safety requirements for the intermediate level liquid waste system

    International Nuclear Information System (INIS)

    1980-07-01

    The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

  15. 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.)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  18. 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

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

    International Nuclear Information System (INIS)

    Golinski, M.; Charomska, K.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Machida, Chuji; Mori, Shoji.

    1976-01-01

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

  1. Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    The mercury-contaminated rinse solution (INEL waste ID number-sign 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 μ to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solut

  2. 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.)

  3. Waste management outlook for mountain regions: Sources and solutions.

    Science.gov (United States)

    Semernya, Larisa; Ramola, Aditi; Alfthan, Björn; Giacovelli, Claudia

    2017-09-01

    Following the release of the global waste management outlook in 2015, the United Nations Environment Programme (UN Environment), through its International Environmental Technology Centre, is elaborating a series of region-specific and thematic waste management outlooks that provide policy recommendations and solutions based on current practices in developing and developed countries. The Waste Management Outlook for Mountain Regions is the first report in this series. Mountain regions present unique challenges to waste management; while remoteness is often associated with costly and difficult transport of waste, the potential impact of waste pollutants is higher owing to the steep terrain and rivers transporting waste downstream. The Outlook shows that waste management in mountain regions is a cross-sectoral issue of global concern that deserves immediate attention. Noting that there is no 'one solution fits all', there is a need for a more landscape-type specific and regional research on waste management, the enhancement of policy and regulatory frameworks, and increased stakeholder engagement and awareness to achieve sustainable waste management in mountain areas. This short communication provides an overview of the key findings of the Outlook and highlights aspects that need further research. These are grouped per source of waste: Mountain communities, tourism, and mining. Issues such as waste crime, plastic pollution, and the linkages between exposure to natural disasters and waste are also presented.

  4. Using bentonite for NPP liquid waste treatment

    International Nuclear Information System (INIS)

    Bui Dang Hanh

    2015-01-01

    During operation, nuclear power plants (NPPs) release a large quantity of water waste containing radionuclides required treatment for protection of the radiation workers and the environment. This paper introduces processes used to treat water waste from Paks NPP in Hungary and it also presents the results of a study on the use of Vietnamese bentonite to remove radioactive Caesium from a simulated water waste containing Cs. (author)

  5. Selectivity of NF membrane for treatment of liquid waste containing uranium

    International Nuclear Information System (INIS)

    Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Afonso, Julio C.

    2013-01-01

    The performance of two nanofiltration membranes were investigated for treatment of liquid waste containing uranium through two conditions permeation: permeation test and concentration test of the waste. In the permeation test solution permeated returned to the feed tank after collected samples each 3 hours. In the test of concentration the permeated was collected continuously until 90% reduction of the feed volume. The liquid waste ('carbonated water') was obtained during conversion of UF 6 to UO 2 in the cycle of nuclear fuel. This waste contains uranium concentration on average 7.0 mg L -1 , and not be eliminated to the environmental. The waste was permeated using a cross-flow membrane cell in the pressure of the 1.5 MPa. The selectivity of the membranes for separation of uranium was between 83% and 90% for both tests. In the concentration tests the waste was concentrated around for 5 times. The surface layer of the membranes was evaluated before and after the tests by infrared spectroscopy (ATR-FTIR), field emission microscopy (FESEM) and atomic force spectroscopy (AFM). The membrane separation process is a technique feasible to and very satisfactory for treatment the liquid waste. (author)

  6. Radioactive waste management at a Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Abrams, C.S.; Fryer, R.H.; Witbeck, L.C.

    1979-01-01

    This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel

  7. Removal of radionuclides from partitioning waste solutions by adsorption and catalytic oxidation methods

    Energy Technology Data Exchange (ETDEWEB)

    Yamagishi, Isao; Yamaguchi, Isoo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kubota, Masumitsu [Research Organization for Information Science and Technology (RIST), Tokai, Ibaraki (Japan)

    2000-09-01

    Adsorption of radionuclides with inorganic ion exchangers and catalytic oxidation of a complexant were studied for the decontamination of waste solutions generated in past partitioning tests with high-level liquid waste. Granulated ferrocyanide and titanic acid were used for adsorption of Cs and Sr, respectively, from an alkaline solution resulting from direct neutralization of an acidic waste solution. Both Na and Ba inhibited adsorption of Sr but Na did not that of Cs. These exchangers adsorbed Cs and Sr at low concentration with distribution coefficients of more than 10{sup 4}ml/g from 2M Na solution of pH11. Overall decontamination factors (DFs) of Cs and total {beta} nuclides exceeded 10{sup 5} and 10{sup 3}, respectively, at the neutralization-adsorption step of actual waste solutions free from a complexant. The DF of total {alpha} nuclides was less than 10{sup 3} for a waste solution containing diethylenetriaminepentaacetic acid (DTPA). DTPA was rapidly oxidized by nitric acid in the presence of a platinum catalyst, and radionuclides were removed as precipitates by neutralization of the resultant solution. The DF of {alpha} nuclides increased to 8x10{sup 4} by addition of the oxidation step. The DFs of Sb and Co were quite low through the adsorption step. A synthesized Ti-base exchanger (PTC) could remove Sb with the DF of more than 4x10{sup 3}. (author)

  8. Study of shrimp shell derivatives for treating of low-level radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Hayeripour, S. [Tonkabon Islamic Azad Univ., Tonkabon (Iran, Islamic Republic of). College of the Environment; Malmasi, S. [North Tehran Islamic Azad Univ., Tehran (Iran, Islamic Republic of). College of the Environment

    2006-07-01

    Chitin derivatives can be used to treat liquid wastes that include heavy metals of radionuclides. In this study, 4 types of chitin derivatives from shrimp shell waste were investigated for their potential in decontaminating and treating low-level radioactive liquid waste (LLW). The adsorption of caesium (Cs); cobalt (Co); and manganese (Mn) isotopes on chitin derivatives were investigated using a batch and column system with variations in diameter, pH, and length of treatment. Chitin derivatives included shrimp shells; de-mineralized shrimp shells; chitin extracted from shrimp shells; and chitosan extracted from shrimp shell waste. Three types of simulated solutions were prepared to study and compare adsorption performance: (1) a mono cationic solution consisting of stable isotopes; (2) a solution containing 3 stable cations; and (3) a simulated radioactive waste containing Cs-137, Co-60, and Mn-54. Results of the experiments showed that all 4 chitin derivatives were capable of adsorbing the isotopes. Despite its low pH, chitosan showed the highest adsorption efficiency. It was concluded that shrimps shells provided unreliable results under different operating conditions. The demineralized shells were suitable for removing Co from solutions. Row shells were not recommended as a suitable adsorbent for radionuclides removal. 14 refs., 2 tabs., 6 figs.

  9. An Ionic Liquid Solution of Chitosan as Organocatalyst

    Directory of Open Access Journals (Sweden)

    René Wilhelm

    2013-11-01

    Full Text Available Chitosan, which is derived from the biopolymer chitin, can be readily dissolved in different ionic liquids. The resulting homogeneous solutions were applied in an asymmetric Aldol reaction. Depending on the type of ionic liquid used, high asymmetric inductions were found. The influence of different additives was also studied. The best results were obtained in [BMIM][Br] without an additive.

  10. Double liquid membrane system for the removal of actinides and lanthanides from acidic nuclear wastes

    International Nuclear Information System (INIS)

    Chiarizia, R.; Danesi, P.R.

    1985-01-01

    Supported liquid membranes (SLM), consisting of an organic solution of n-octyl-(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl-phosphate (TBP) in decalin are able to perform selective separation and concentration of actinide and lanthanide ions from aqueous nitrate feed solutions and synthetic nuclear wastes. In the membrane process a possible strip solution is a mixture of formic acid and hydroxylammonium formate (HAF). The effectiveness of this strip solution is reduced and eventually nullified by the simultaneous transfer through the SLM of nitric acid which accumulates in the strip solution. A possible way to overcome this drawback is to make use of a second SLM consisting of a primary amine which is able to extract only HNO 3 from the strip solution. In this work the results obtained by experimentally studying the membrane system: synthetic nuclear waste/CMPO-TBP membrane/HCOOH-HAF strip solution/primary amine membrane/NaOH solution, are reported. They show that the use of a second liquid membrane is effective in controlling the HNO 3 concentration in the strip solution, thus allowing the actinide and lanthanide ions removal from the feed solution to proceed to completion. 15 refs., 10 figs., 1 tab

  11. Experiments on the possible usage of liquid industrial wastes from a paint and lacquer factory for flue gas desulphurization

    Energy Technology Data Exchange (ETDEWEB)

    Trzepierczynska, I.; Lech-Brzyk, K. [Technical University of Wroclaw, Wroclaw (Poland). Inst. of Environment Protection Engineering

    1995-12-31

    In this paper, the complex solution of environment protection against flue gases (comprising sulphur dioxide) and alkaline industrial wastewater is provided. Industrial wastes from a paint and lacquer factory were examined and their usage for sulphur dioxide absorption was determined. The combined method of alkaline waste neutralization and flue gas desulphurization is proposed. The liquid wastes come from the POLIFARB SA plant in Wroclaw. 9 refs., 7 tabs.

  12. Assessment and analysis of industrial liquid waste and sludge disposal at unlined landfill sites in arid climate

    International Nuclear Information System (INIS)

    Al Yaqout, Anwar F.

    2003-01-01

    Municipal solid waste disposal sites in arid countries such as Kuwait receive various types of waste materials like sewage sludge, chemical waste and other debris. Large amounts of leachate are expected to be generated due to the improper disposal of industrial wastewater, sewage sludge and chemical wastes with municipal solid waste at landfill sites even though the rainwater is scarce. Almost 95% of all solid waste generated in Kuwait during the last 10 years was dumped in five unlined landfills. The sites accepting liquid waste consist of old sand quarries that do not follow any specific engineering guidelines. With the current practice, contamination of the ground water table is possible due to the close location of the water table beneath the bottom of the waste disposal sites. This study determined the percentage of industrial liquid waste and sludge of the total waste dumped at the landfill sites, analyzed the chemical characteristics of liquid waste stream and contaminated water at disposal sites, and finally evaluated the possible risk posed by the continuous dumping of such wastes at the unlined landfills. Statistical analysis has been performed on the disposal and characterization of industrial wastewater and sludge at five active landfill sites. The chemical analysis shows that all the industrial wastes and sludge have high concentrations of COD, suspended solids, and heavy metals. Results show that from 1993 to 2000, 5.14±1.13 million t of total wastes were disposed per year in all active landfill sites in Kuwait. The share of industrial liquid and sludge waste was 1.85±0.19 million t representing 37.22±6.85% of total waste disposed in all landfill sites. Such wastes contribute to landfill leachate which pollutes groundwater and may enter the food chain causing adverse health effects. Lined evaporation ponds are suggested as an economical and safe solution for industrial wastewater and sludge disposal in the arid climate of Kuwait

  13. Electrochemical processing of nitrate waste solutions

    International Nuclear Information System (INIS)

    Genders, D.; Weinberg, N.; Hartsough, D.

    1992-01-01

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F - ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions

  14. 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

  15. Incineration plant for thermal destruction of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Bartoli, B.; Lisbonne, P.

    1988-01-01

    Incineration was selected to destroy organic liquids contaminated by radioelements. This treatment offers the advantage of reducing the volume of wastes considerably. Therefore an incineration plant has been built within the nuclear research center of Cadarache. After an experimental work with inactive organic liquids from June 1980 to March 1981, the incineration plant was approved by safety authorities for the incineration of contaminated organic liquids. The capacity ranges from 20l/hr to 50l/hr. On the basis of 6 years of operation and a volume of 200 m3 the incineration plant has shown reliable operating conditions in the destruction of various contaminated organic liquids

  16. The best solution to our Nation's waste management problem: Education

    International Nuclear Information System (INIS)

    Mikel, C.J.

    1992-01-01

    In addition to the Waste Isolation Pilot Plant (WIPP) being the best solution today to the Nation's problem of permanent storage of transuranic radioactive waste produced by the defense industry, WIPP is also involved in finding the solution for another national problem: the education of our youth. The youth of America have grown up thinking that science and math are too hard, or not interesting. We, the parents of our Nation's leaders of tomorrow, must find a solution to this dilemma. It is the mission of the Waste Isolation Division Educational Programs to create programs to promote quality education in the classroom and to enhance each student's interest in mathematics and the sciences

  17. Combustion of animal or vegetable based liquid waste products

    International Nuclear Information System (INIS)

    Wikman, Karin; Berg, Magnus

    2002-04-01

    In this project experiences from combustion of animal and vegetable based liquid waste products have been compiled. Legal aspects have also been taken into consideration and the potential for this type of fuel on the Swedish energy market has been evaluated. Today the supply of animal and vegetable based liquid waste products for energy production in Sweden is limited. The total production of animal based liquid fat is about 10,000 tonnes annually. The animal based liquid waste products origin mainly from the manufacturing of meat and bone meal. Since meat and bone meal has been banned from use in animal feeds it is possible that the amount of animal based liquid fat will decrease. The vegetable based liquid waste products that are produced in the processing of vegetable fats are today used mainly for internal energy production. This result in limited availability on the commercial market. The potential for import of animal and vegetable based liquid waste products is estimated to be relatively large since the production of this type of waste products is larger in many other countries compared to Sweden. Vegetable oils that are used as food or raw material in industries could also be imported for combustion, but this is not reasonable today since the energy prices are relatively low. Restrictions allow import of SRM exclusively from Denmark. This is today the only limit for increased imports of animal based liquid fat. The restrictions for handle and combustion of animal and vegetable based liquid waste products are partly unclear since this is covered in several regulations that are not easy to interpret. The new directive for combustion of waste (2000/76/EG) is valid for animal based waste products but not for cadaver or vegetable based waste products from provisions industries. This study has shown that more than 27,400 tonnes of animal based liquid waste products and about 6,000 tonnes of vegetable based liquid waste products were used for combustion in Sweden

  18. 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.)

  19. Expert system for liquid low-level waste management

    International Nuclear Information System (INIS)

    Ferrada, J.J.

    1992-01-01

    An expert system prototype has been developed to support system analysis activities at the Oak Ridge National Laboratory (ORNL) for waste management tasks. This expert system will aid in prioritizing radioactive waste streams for treatment and disposal by evaluating the severity and treatability of the problem as well as the final waste form. The objectives of the expert system development included: (1) collecting information on process treatment technologies for liquid low-level waste (LLLW) that can be incorporated in the knowledge base of the expert system, and (2) producing a prototype that suggests processes and disposal technologies for the ORNL LLLW system. The concept under which the expert system has been designed is integration of knowledge. There are many sources of knowledge (data bases, text files, simulation programs, etc.) that an expert would regularly consult in order to solve a problem of liquid waste management. The expert would normally know how to extract the information from these different sources of knowledge. The general scope of this project would be to include as much pertinent information as possible within the boundaries of the expert system. As a result, the user, who may not be an expert in every aspect of liquid waste management, may be able to apply the content of the information to a specific waste problem. This paper gives the methodological steps to develop the expert system under this general framework

  20. Treatment of Industrial Liquid Waste of Steel Plating by Coagulation-Flocculation Using Sodium Biphosphate

    International Nuclear Information System (INIS)

    Subiarto; Herlan Martono

    2007-01-01

    Research about treatment of industrial liquid waste of steel plating by coagulation-flocculation using sodium biphosphate have been conducted. The purpose of the treatment was the content reduction of Cr, Ni, and Cu in the liquid waste, so that produced effluent with Cr, Ni, and Cu content until they laid under mutual standard. The variables studied in this process were the solution pH, the coagulant/waste volume comparison, the speed of the fast stirring, and the time of the fast stirring. Optimum separation efficiency on coagulation-flocculation process of liquid waste of steel plating using sodium biphosphate at the condition of solution ph 9, coagulant/waste volume comparation 1.50, the speed of the fast stirring 400 rpm, and the time of fast stirring is 5 minute. Low stirring was conducted at 60 rpm for 60 minute. The yields of optimum separation efficiency in this condition were 99.48 % for Cr, 99.51 % for Ni, and 99.03 % for Cu. (author)

  1. Recovery of Ionic Liquids from aqueous solution by Nanofiltration

    OpenAIRE

    Fernández Dámaso, José Francisco

    2011-01-01

    The T-SAR methodology was combined with membrane characterization methods. An application of the combined approach was demonstrated with two commercial nanofiltration membranes and it was possible to successfully predict their performance for the recovery of ionic liquids from aqueous solution. Using model solutions of Pyr16 (CF3SO2)2N, it could be evidenced the formation of a new phase of ionic liquid during the concentration process. In this case, 66% of the ionic liquid was separated and t...

  2. Glass-solidification method for high level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

    1996-01-01

    High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

  3. Importance of waste composition for Life Cycle Assessment of waste management solutions

    DEFF Research Database (Denmark)

    Bisinella, Valentina; Götze, Ramona; Conradsen, Knut

    2017-01-01

    The composition of waste materials has fundamental influence on environmental emissions associated with waste treatment, recycling and disposal, and may play an important role also for the Life Cycle Assessment (LCA) of waste management solutions. However, very few assessments include effects...... of the waste composition and waste LCAs often rely on poorly justified data from secondary sources. This study systematically quantifiesy the influence and uncertainty on LCA results associated with selection of waste composition data. Three archetypal waste management scenarios were modelled with the waste...... LCA model EASETECH based on detailed waste composition data from the literature. The influence from waste composition data on the LCA results was quantified with a step-wise Global Sensitivity Analysis (GSA) approach involving contribution, sensitivity, uncertainty and discernibility analyses...

  4. Deep injection disposal of liquid radioactive waste in Russia

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Acid fractionation for low level liquid waste cleanup and recycle

    International Nuclear Information System (INIS)

    Gombert, D. II; McIntyre, C.V.; Mizia, R.E.; Schindler, R.E.

    1990-01-01

    At the Idaho Chemical Processing Plant, low level liquid wastes containing small amounts of radionuclides are concentrated via a thermosyphon evaporator for calcination with high level waste, and the evaporator condensates are discharged with other plant wastewater to a percolation pond. Although all existing discharge guidelines are currently met, work has been done to reduce all waste water discharges to an absolute minimum. In this regard, a 15-tray acid fractionation column will be used to distill the mildly acidic evaporator condensates into concentrated nitric acid for recycle in the plant. The innocuous overheads from the fractionator having a pH greater than 2, are superheated and HEPA filtered for atmospheric discharge. Nonvolatile radionuclides are below detection limits. Recycle of the acid not only displaces fresh reagent, but reduces nitrate burden to the environment, and completely eliminates routine discharge of low level liquid wastes to the environment

  6. Technical solutions for waste treatment in the Belene project

    International Nuclear Information System (INIS)

    Büttner, K.; Eichhorn, H.

    2011-01-01

    Outline: In June 2010 NUKEM Technologies GmbH was awarded a contract from ATOMSTROYEXPORT JSC to perform the complete work package related to designing and completion of the equipment for treatment of radioactive waste on the turn-key basis for Belene NPP. Technical Solutions: Waste Streams and Technologies at UKC and UKS; Concentration Plant; Thermal Treatment of Resins Sorting Facility; Biological Waste Water Treatment; Conditioning – Cementation • Sorting of Radwaste; Plasma Facility; Grouting; Filter Press; Monitoring and Tracking

  7. Electrochemical processing of low-level waste solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Ebra, M.A.

    1987-01-01

    The feasibility of treating low-level Savannah River Plant (SRP) waste solutions by an electrolytic process has been demonstrated. Although the economics of the process are marginal at the current densities investigated at the laboratory scale, there are a number of positive environmental benefits. These benefits include: (1) reduction in the levels of nitrate and nitrite in the waste, (2) further decontamination of 99 Tc and 106 Ru, and (3) reduction in the volume of waste

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  10. Industrial Water Waste, Problems and the Solution

    Directory of Open Access Journals (Sweden)

    Alif Noor Anna

    2004-01-01

    Full Text Available Recently, the long term development in Indonesia has changed agricultural sector to the industrial sector. This development can apparently harm our own people. This is due to the waste that is produced from factories. The waste from various factories seems to have different characteristics. This defference encourages us to be able to find out different of methods of managing waste so that cost can be reduced, especially in water treatment. In order that industrial development and environmental preservation can run together in balance, many institutions involved should be consider, especially in the industrial chain, the environment, and human resource, these three elements can be examined in terms of their tolerance to waste.

  11. Method of processing radioactive liquid wastes by using zeolites

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, T; Mimura, H

    1975-09-18

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

  12. Technical report on treatment of radioactive slurry liquid waste

    International Nuclear Information System (INIS)

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

    1999-06-01

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

  13. 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.

  14. Environment friendly solutions of plastics waste management

    International Nuclear Information System (INIS)

    Pirzada, F.N.; Riffat, T.; Pirzada, M.D.S.

    1997-01-01

    The use of plastics is growing worldwide. Consequently, the volume of plastic waste is also increasing. Presently, more than 100 million tons per year of plastic is being produced globally. In U.S. alone more than 10 million tons of plastic is being dumped in landfills as waste, where it can persist for decades. This has resulted in exhausting old landfills. Public awareness on environment is also making it difficult to find new sites for landfills. This has led to increased emphasis on treatment and recycling of plastic wastes. Volume reduction of plastic waste has some unique problems. They arise from the intrinsic chemical inertness of polymeric materials and toxic nature of their degradation byproducts. The paper reviews the present state of plastic waste management including land filling, incineration and recycling technologies. The technical problems associated with each of these processes have been discussed. There is also brief description of ongoing R and D for finding improved methods of plastic waste handling with their promises and problems. The role of tougher legislation in developing better recycling methods and degradable plastics has also been evaluated. The claims made by the proponents of degradable polymers have also been critically reviewed. (authors)

  15. 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)

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

    International Nuclear Information System (INIS)

    Irie, Hiromitsu; Tajima, Fumio.

    1975-01-01

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

  17. Solidification of intermediate level liquid waste - ILLW, CEMEX waste form qualification

    International Nuclear Information System (INIS)

    D'Andrea, V.; Guerra, M.; Pancotti, F.; Maio, V.

    2015-01-01

    In the Sogin EUREX Facility about 125 m 3 of intermediate level radioactive waste and about 113 m 3 of low level radioactive waste, produced during the re-processing of MTR and CANDU fuel, are stored. Solidification of these wastes is planned in order to fulfill the specific requirements established by the Safety Authority, taking into account the criteria set up in a Technical Guide on the issue of radioactive waste management. The design of a cementation plant (CEMEX) of all liquid radioactive wastes is currently ongoing. The process requires that the liquid waste is neutralized with NaOH (NaOH 19 M) and metered into 440 liter drum together with the cement, while the mixture is stirred by a lost paddle ('in drum mixing process'). The qualification of the Waste Form consists of all the activities demonstrating that the final cemented product has the minimum requirements (mechanical, chemical and physical characteristics) compliant with all the subsequent management phases: long-term interim storage, transport and long-term disposal of the waste. All tests performed to qualify the conditioning process for immobilizing first extraction cycle (MTR and CANDU) and second extraction cycle liquid wastes, gave results in compliance with the minimum requirements established for disposal

  18. Preliminary study for treatment methodology establishment of liquid waste containing uranium in refining facility lagoon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Jik; Lee, Kune Woo; Won, Hui Jun; Ahn, Byung Gil; Shim, Joon Bo

    1999-12-01

    The preliminary study which establishes the treatment methodology of the sludge waste containing uranium in the conversion facility lagoon was performed. The property of lagoon liquid waste such as the initial water content, the density including radiochemical analysis results were obtained using the samples taken from the lagoon. The objective of this study is to provide some basically needed materials for selection of the most proper lagoon waste treatment methodology by reviewing the effective processes and methods for minimizing the secondary waste resulting from the treatment and disposition of large amount of radioactive liquid waste according to the facility closing. The lagoon waste can be classified into two sorts, such as supernatant and precipitate. The supernatants contain uranium less than 5 ppm and their water content are about 35 percent. Therefore, supernatants are solutions composed of mainly salt components. However, the precipitates have lots of uranium compound contained in the coagulation matrix, and are formed as two kinds of crystalline structures. The most proper method minimizing the secondary waste would be direct drying and solidification of the supernatants and precipitates after separation of them by filtering. (author)

  19. Removal of dissolved and suspended radionuclides from Hanford Waste Vitrification Plant liquid wastes

    International Nuclear Information System (INIS)

    Sharp, S.D.; Nankani, F.D.; Bray, L.A.; Eakin, D.E.; Larson, D.E.

    1990-12-01

    It was determined during Preliminary Design of the Hanford Waste Vitrification Plant that certain intermediate process liquid waste streams should be decontaminated in a way that would permit the purge of dissolved chemical species from the process recycle shop. This capability is needed to ensure proper control of product glass chemical composition and to avoid excessive corrosion of process equipment. This paper discusses the process design of a system that will remove both radioactive particulates and certain dissolved fission products from process liquid waste streams. Supporting data obtained from literature sources as well as from laboratory- and pilot-scale tests are presented. 3 refs., 1 fig., 3 tabs

  20. Liquid and Gaseous Waste Operations Department Annual Operating Report, CY 1993

    International Nuclear Information System (INIS)

    Maddox, J.J.; Scott, C.B.

    1994-02-01

    This report summarizes the activities of the waste management operations section of the liquid and gaseous waste operations department at ORNL for 1993. The process waste, liquid low-level waste, gaseous waste systems activities are reported, as well as the low-level waste solidification project. Upgrade activities is the various waste processing and treatment systems are summarized. A maintenance activity overview is provided, and program management, training, and other miscellaneous activities are covered

  1. Treatment alternatives of liquid radioactive waste containing uranium in phosphoric acid

    International Nuclear Information System (INIS)

    Bustamante Escobedo, Mauricio

    2003-01-01

    The UGDR, receives annually 100 [l] of liquid radioactive waste containing, highly acid (pH=0) uranium in phosphoric acid from the Laboratory of Chemical Analysis. This waste must be chemically and radiologically decontaminated before it can be discharged in accordance with local environmental standards. Chemical precipitation and evaporation test were carried out to define the operating conditions for the radiological decontamination of this radioactive waste and to obtain a solid waste that can be conditioned in a cement matrix. The evaporation process generates excellent rates of volume reduction, over 80%, but generates a pulp that is hard handle when submitted to a drying process. Chemical precipitation generates good results for decontaminating these solutions and reducing volume (above 50%) to obtain a uranium free effluent. The treatment with calcium carbonate generated an effluent with a low concentration of polluting agents. A preliminary test was carried out condition these solids in a cement matrix, using ratios of 0.45 waste/cement and 2 of water/cement. The mix prepared with waste from the sodium hydroxide treatment had low mechanical resistance resulting from the saline incrustations. The waste from the calcium carbonate treatment was very porous due to the water evaporation from the highly exothermic reaction between the waste and the cement. The mix of the calcium carbonate generated waste and the cement matrix needs to be optimized, since it generates favorable conditions for adhering with the cement matrix (au)

  2. 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)

  3. SPEEDUP simulation of liquid waste batch processing. Revision 1

    International Nuclear Information System (INIS)

    Shannahan, K.L.; Aull, J.E.; Dimenna, R.A.

    1994-01-01

    The Savannah River Site (SRS) has accumulated radioactive hazardous waste for over 40 years during the time SRS made nuclear materials for the United States Department of Energy (DOE) and its predecessors. This waste is being stored as caustic slurry in a large number of 1 million gallon steel tanks, some of which were initially constructed in the early 1950's. SRS and DOE intend to clean up the Site and convert this waste into stable forms which then can be safely stored. The liquid waste will be separated into a partially decontaminated low-level and radioactive high-level waste in one feed preparation operation, In-Tank Precipitation. The low-level waste will be used to make a concrete product called saltstone in the Saltstone Facility, a part of the Defense Waste Processing Facility (DWPF). The concrete will be poured into large vaults, where it will be permanently stored. The high-level waste will be added to glass-formers and waste slurry solids from another feed preparation operation, Extended Sludge Processing. The mixture will then be converted to a stable borosilicate glass by a vitrification process that is the other major part of the DWPF. This glass will be poured into stainless steel canisters and sent to a temporary storage facility prior to delivery to a permanent underground storage site

  4. Modified microspheres for cleaning liquid wastes from radioactive nuclides

    International Nuclear Information System (INIS)

    Danilin, Lev; Drozhzhin, Valery

    2007-01-01

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

  5. effect of municipal liquid waste on corrosion susceptibility

    African Journals Online (AJOL)

    DR. AMINU

    Kogo, A. A.. Department of Integrated Science, Federal College of Education, Kano, Nigeria. ... The corrosion rate of the galvanized steel pipe was measured using the gravimetric ... Key words: Liquid waste, galvanized steel, weight loss, gravimetric, corrosion, leaking ... the side of the test tubes, so that each side would be.

  6. Liquid waste management: The case of Bahir Dar, Ethiopia

    African Journals Online (AJOL)

    admin

    liquid waste management practices of the community; to assess the .... Logistic regression was performed to assess the impact of a number of factors on the .... the ever-growing Bahir Dar Town with modern buildings using flush toilets will ...

  7. Device for the disposal of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Tomizawa, Toshi; Inoue, Tadashi.

    1976-01-01

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

  8. Liquid waste management at nuclear power plant with WWER

    International Nuclear Information System (INIS)

    Sabouni, Zahra.

    1995-07-01

    Management of radioactive wastes have become an area of ever increasing important in nuclear power plants. This is due to the fact that national and international regulations will only allow activity release to the environment based on ALARA principles. Radioactive liquids in the nuclear power plant originate as leakage from equipment, as drains from reactor and auxiliary systems, from decontamination and cleaning operations, from active laundry and from personnel showers. They will collected through the controlled zone of the plant in sumps and automatically pumped to large tanks and then to treatment system. The radioactive wastes are separated and categorized according to their main physical and chemical properties. Methods most frequently applied for low and intermediate level; liquid wastes are: chemical treatment (precipitation), ion exchange, and evaporation, and the decontamination ors are a few hundred, 10 2 -10 4 and 10 3 -10 6 , respectively. As a result of the treatment of radioactive liquids by mentioned methods a concentration of activity takes place in filter media, ion exchange resins, and evaporator concentrates. Before the semi-solid wastes shipped for storage, it has to be solidified in order to handle and transport in easier way. The solidification of wastes can take place by different methods. The general methods are: cementation, and bituminization processes. The selection of each process will depend on many factors which should be considered during the design phase. (author)

  9. Purification process for aqueous solutions of rare earths by liquid-liquid extraction

    International Nuclear Information System (INIS)

    Rollat, A.; Sabot, J.L.; Burgard, M.; Delloye, T.

    1986-01-01

    Alkaline earth metals are removed by liquid-liquid extraction between on aqueous nitric phase of impure rare earth compounds and an organic phase of polyether (crown ether). This process is particularly suited to removal of Ca, Ba and Ra contained in nitric solutions of rare earths [fr

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

    KAUST Repository

    Lee, Songbok

    2018-04-07

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

  11. Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

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

    International Nuclear Information System (INIS)

    Yossef, Y.A.A.

    2013-01-01

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

  13. Engineering solutions to the management of solid radioactive waste

    International Nuclear Information System (INIS)

    1991-01-01

    The management of radioactive waste, its safe handling and ultimate disposal, is of vital concern to engineers in the nuclear industry. The international conference 'Engineering Solutions to the Management of Solid Radioactive Waste', organized by the Institution of Mechanical Engineers and held in Manchester in November 1991, provided a forum for the discussion and comparison of the different methods of waste management used in Europe and America. Papers presented and discussed included: the interaction between the design of containers for low level radioactive waste and the design of a deep repository, commercial low level waste disposal sites in the United States, and the development of radioactive waste monitoring systems at the Sellafield reprocessing complex. This volume is a collection of 22 papers presented at the conference. All are indexed separately. (author)

  14. Separation of actinides and lanthanides from acidic nuclear wastes by supported liquid membranes

    International Nuclear Information System (INIS)

    Danesi, P.R.; Chiarizia, R.; Rickert, P.; Horwitz, E.P.

    1985-01-01

    Supported liquid membranes, SLM, consisting of a solution of 0.25 M octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and 0.75 M tributylphosphate (TBP) in decalin absorbed on thin microporous polypropylene supports, have been studied for their ability to perform selective separations and concentrations of actinide and lanthanide ions from synthetic acidic nuclear wastes. The permeability coefficients of selected actinides (Am, Pu, U, Np) and of some of the other major components of the wastes have been measured using SLMs in flat-sheet and hollow-fiber configurations. The results have shown that with the thin (25 μm) flat-sheet SLMs, using Celgard 2500 as support, the membrane permeation process is mainly controlled by the rate of diffusion through the aqueous boundary layers. With the thicker (430 μm) hollow-fiber SLMs, using Accurel hollow-fibers as support, the membrane permeation process is controlled by the rate of diffusion through both the SLM and the aqueous boundary layers. Hollow-fibers SLMs exhibited lower permeability coefficients and longer life-times. The experiments have shown that the actinides can be very efficiently removed from the synthetic waste solutions to the point that the resulting solution could be considered a non-transuranic waste (less than 100 mCi/g of disposed form). The work has demonstrated that actinide removal from synthetic waste solutions is a feasible chemical process at the laboratory scale level

  15. 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 )

  16. 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.

  17. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)

    2014-09-15

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  18. 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

  19. Calcium carbonate synthesis with prescribed properties based on liquid waste of soda production

    Directory of Open Access Journals (Sweden)

    E.O. Mikhailova

    2016-09-01

    Full Text Available A promising direction in solving of environmental problems of soda industry is the development of low-waste resource-saving technologies, which consist in recycling of valuable waste components with obtaining the commercial products. Aim: The aim is to establish the optimal conditions for obtaining calcium carbonate with prescribed properties from liquid waste of soda production. Materials and Methods: Chemically deposited calcium carbonate is used as filler and should have certain physical and chemical properties. To obtain a product of prescribed quality the process of calcium carbonate deposition was performed of still waste liquid, that is the waste of calcium carbonate production and contain significant amount of calcium ions, and excessive production of the purified stock solution of sodium bicarbonate, which is composed of carbonate and hydrocarbonate ions. Results: The dependence of bulk density and specific surface area of calcium carbonate sediments and degree of deposition from such technological parameters are established: method of mixing the stock solutions, the concentration and molar ratio of reactants, temperature and reaction time. Conclusions: The optimal mode of deposition process is determined and the concept of production of calcium carbonate is developed. The quality of calcium carbonate meets the modern requirements of high dispersion, low bulk density and evolved specific surface of the product.

  20. Soft X-ray Absorption Spectroscopy of Liquids and Solutions.

    Science.gov (United States)

    Smith, Jacob W; Saykally, Richard J

    2017-12-13

    X-ray absorption spectroscopy (XAS) is an electronic absorption technique for which the initial state is a deeply buried core level. The photon energies corresponding to such transitions are governed primarily by the binding energies of the initial state. Because the binding energies of core electrons vary significantly among atomic species, this makes XAS an element-selective spectroscopy. Proper interpretation of XA spectra can provide detailed information on the local chemical and geometric environment of the target atom. The introduction of liquid microjet and flow cell technologies into XAS experiments has enabled the general study of liquid samples. Liquids studied to date include water, alcohols, and solutions with relevance to biology and energy technology. This Review summarizes the experimental techniques employed in XAS studies of liquid samples and computational methods used for interpretation of the resulting spectra and summarizes salient experiments and results obtained in the XAS investigations of liquids.

  1. Sustainable solutions for solid waste management in Southeast Asian countries

    International Nuclear Information System (INIS)

    Uyen Nguyen Ngoc; Schnitzer, Hans

    2009-01-01

    Human activities generate waste and the amounts tend to increase as the demand for quality of life increases. Today's rate in the Southeast Asian Nations (ASEANs) is alarming, posing a challenge to governments regarding environmental pollution in the recent years. The expectation is that eventually waste treatment and waste prevention approaches will develop towards sustainable waste management solutions. This expectation is for instance reflected in the term 'zero emission systems'. The concept of zero emissions can be applied successfully with today's technical possibilities in the agro-based processing industry. First, the state-of-the-art of waste management in Southeast Asian countries will be outlined in this paper, followed by waste generation rates, sources, and composition, as well as future trends of waste. Further on, solutions for solid waste management will be reviewed in the discussions of sustainable waste management. The paper emphasizes the concept of waste prevention through utilization of all wastes as process inputs, leading to the possibility of creating an ecosystem in a loop of materials. Also, a case study, focusing on the citrus processing industry, is displayed to illustrate the application of the aggregated material input-output model in a widespread processing industry in ASEAN. The model can be shown as a closed cluster, which permits an identification of opportunities for reducing environmental impacts at the process level in the food processing industry. Throughout the discussion in this paper, the utilization of renewable energy and economic aspects are considered to adapt to environmental and economic issues and the aim of eco-efficiency. Additionally, the opportunities and constraints of waste management will be discussed.

  2. 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

  3. 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)

  4. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  5. Solutions of group IV elements in liquid lithium

    International Nuclear Information System (INIS)

    Dadd, A.T.; Hubberstey, P.; Roberts, P.G.

    1982-01-01

    The solubilities of tin (0.00 = 22 Sn 5 . A simple thermochemical cycle is used to demonstrate that, whereas carbon dissolves endothermically in both liquid lithium and liquid sodium, the heavier Group IV elements dissolve exothermically. A similar cycle is used to derive solvation enthalpies (for the neutral gaseous species) for all Group IV elements in the two solvents. The trend in solvation enthalpy: C > Si > Ge > Sn > Pb is indicative of a diminishing affinity of solvent for solute and is attributed to the increasing metallic character of the solute as the Group is descended. (author)

  6. 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.

  7. 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

  8. 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

  9. Turbulence effects on volatilization rates of liquids and solutes

    Science.gov (United States)

    Lee, J.-F.; Chao, H.-P.; Chiou, C.T.; Manes, M.

    2004-01-01

    Volatilization rates of neat liquids (benzene, toluene, fluorobenzene, bromobenzene, ethylbenzene, m-xylene, o-xylene, o-dichlorobenzene, and 1-methylnaphthalene) and of solutes (phenol, m-cresol, benzene, toluene, ethylbenzene, o-xylene, and ethylene dibromide) from dilute water solutions have been measured in the laboratory over a wide range of air speeds and water-stirring rates. The overall transfer coefficients (KL) for individual solutes are independent of whether they are in single- or multi-solute solutions. The gas-film transfer coefficients (kG) for solutes in the two-film model, which have hitherto been estimated by extrapolation from reference coefficients, can now be determined directly from the volatilization rates of neatliquids through anew algorithm. The associated liquid-film transfer coefficients (KL) can then be obtained from measured KL and kG values and solute Henry law constants (H). This approach provides a novel means for checking the precision of any kL and kG estimation methods for ultimate prediction of KL. The improved kG estimation enables accurate K L predictions for low-volatility (i.e., low-H) solutes where K L and kGH are essentially equal. In addition, the prediction of KL values for high-volatility (i.e., high-H) solutes, where KL ??? kL, is also improved by using appropriate reference kL values.

  10. Evaluation of interim and final waste forms for the newly generated liquid low-level waste flowsheet

    International Nuclear Information System (INIS)

    Abotsi, G.M.K.; Bostick, D.T.; Beck, D.E.

    1996-05-01

    The purpose of this review is to evaluate the final forms that have been proposed for radioactive-containing solid wastes and to determine their application to the solid wastes that will result from the treatment of newly generated liquid low-level waste (NGLLLW) and Melton Valley Storage Tank (MVST) supernate at the Oak Ridge National Laboratory (ORNL). Since cesium and strontium are the predominant radionuclides in NGLLLW and MVST supernate, this review is focused on the stabilization and solidification of solid wastes containing these radionuclides in cement, glass, and polymeric materials-the principal waste forms that have been tested with these types of wastes. Several studies have shown that both cesium and strontium are leached by distilled water from solidified cement, although the leachabilities of cesium are generally higher than those of strontium under similar conditions. The situation is exacerbated by the presence of sulfates in the solution, as manifested by cracking of the grout. Additives such as bentonite, blast-furnace slag, fly ash, montmorillonite, pottery clay, silica, and zeolites generally decrease the cesium and strontium release rates. Longer cement curing times (>28 d) and high ionic strengths of the leachates, such as those that occur in seawater, also decrease the leach rates of these radionuclides. Lower cesium leach rates are observed from vitrified wastes than from grout waste forms. However, significant quantities of cesium are volatilized due to the elevated temperatures required to vitrify the waste. Hence, vitrification will generally require the use of cleanup systems for the off-gases to prevent their release into the atmosphere

  11. Fabrication of Polyacrylonitrile Hollow Fiber Membranes from Ionic Liquid Solutions

    KAUST Repository

    Kim, Dooli; Moreno Chaparro, Nicolas; Nunes, Suzana Pereira

    2015-01-01

    The interest in green processes and products has increased to reduce the negative impact of many industrial processes to the environment. Solvents, which play a crucial role in the fabrication of membranes, need to be replaced by sustainable and less toxic solvent alternatives for commonly used polymers. The purpose of this study is the fabrication of greener hollow fiber membranes based on polyacrylonitrile (PAN), substituting dimethylformamide (DMF) by less toxic mixtures of ionic liquids (IL) and dimethylsulfoxide (DMSO). A thermodynamic analysis was conducted, estimating the Gibbs free energy of mixing to find the most convenient solution compositions. Hollow fiber membranes were manufactured and optimized. As a result, a uniform pattern and high porosity were observed in the inner surface of the membranes prepared from the ionic liquid solutions. The membranes were coated with a polyamide layer by interfacial polymerization the hollow fiber membranes were applied in forward osmosis experiments by using sucrose solutions as draw solution.

  12. Fabrication of Polyacrylonitrile Hollow Fiber Membranes from Ionic Liquid Solutions

    KAUST Repository

    Kim, Dooli

    2015-10-08

    The interest in green processes and products has increased to reduce the negative impact of many industrial processes to the environment. Solvents, which play a crucial role in the fabrication of membranes, need to be replaced by sustainable and less toxic solvent alternatives for commonly used polymers. The purpose of this study is the fabrication of greener hollow fiber membranes based on polyacrylonitrile (PAN), substituting dimethylformamide (DMF) by less toxic mixtures of ionic liquids (IL) and dimethylsulfoxide (DMSO). A thermodynamic analysis was conducted, estimating the Gibbs free energy of mixing to find the most convenient solution compositions. Hollow fiber membranes were manufactured and optimized. As a result, a uniform pattern and high porosity were observed in the inner surface of the membranes prepared from the ionic liquid solutions. The membranes were coated with a polyamide layer by interfacial polymerization the hollow fiber membranes were applied in forward osmosis experiments by using sucrose solutions as draw solution.

  13. WASTE TREATMENT PLANT (WTP) LIQUID EFFLUENT TREATABILITY EVALUATION

    International Nuclear Information System (INIS)

    LUECK, K.J.

    2004-01-01

    A forecast of the radioactive, dangerous liquid effluents expected to be produced by the Waste Treatment Plant (WTP) was provided by Bechtel National, Inc. (BNI 2004). The forecast represents the liquid effluents generated from the processing of Tank Farm waste through the end-of-mission for the WTP. The WTP forecast is provided in the Appendices. The WTP liquid effluents will be stored, treated, and disposed of in the Liquid Effluent Retention Facility (LERF) and the Effluent Treatment Facility (ETF). Both facilities are located in the 200 East Area and are operated by Fluor Hanford, Inc. (FH) for the US. Department of Energy (DOE). The treatability of the WTP liquid effluents in the LERF/ETF was evaluated. The evaluation was conducted by comparing the forecast to the LERF/ETF treatability envelope (Aromi 1997), which provides information on the items which determine if a liquid effluent is acceptable for receipt and treatment at the LERF/ETF. The format of the evaluation corresponds directly to the outline of the treatability envelope document. Except where noted, the maximum annual average concentrations over the range of the 27 year forecast was evaluated against the treatability envelope. This is an acceptable approach because the volume capacity in the LERF Basin will equalize the minimum and maximum peaks. Background information on the LERF/ETF design basis is provided in the treatability envelope document

  14. Chromium liquid waste inertization in an inorganic alkali activated matrix: Leaching and NMR multinuclear approach

    International Nuclear Information System (INIS)

    Ponzoni, Chiara; Lancellotti, Isabella; Barbieri, Luisa; Spinella, Alberto; Saladino, Maria Luisa; Martino, Delia Chillura; Caponetti, Eugenio; Armetta, Francesco; Leonelli, Cristina

    2015-01-01

    Highlights: • Inertization of chromium liquid waste in aluminosilicate matrix. • Water less inertization technique exploiting the waste water content. • Liquid waste inertization without drying step. • Long term stabilization study through leaching test. • SEM analysis and 29 Si and 27 Al MAS NMR in relation with long curing time. - Abstract: A class of inorganic binders, also known as geopolymers, can be obtained by alkali activation of aluminosilicate powders at room temperature. The process is affected by many parameters (curing time, curing temperature, relative humidity etc.) and leads to a resistant matrix usable for inertization of hazardous waste. In this study an industrial liquid waste containing a high amount of chromium (≈2.3 wt%) in the form of metalorganic salts is inertized into a metakaolin based geopolymer matrix. One of the innovative aspects is the exploitation of the water contained in the waste for the geopolymerization process. This avoided any drying treatment, a common step in the management of liquid hazardous waste. The evolution of the process - from the precursor dissolution to the final geopolymer matrix hardening - of different geopolymers containing a waste amount ranging from 3 to 20% wt and their capability to inertize chromium cations were studied by: i) the leaching tests, according to the EN 12,457 regulation, at different curing times (15, 28, 90 and 540 days) monitoring releases of chromium ions (Cr(III) and Cr(VI)) and the cations constituting the aluminosilicate matrix (Na, Si, Al); ii) the humidity variation for different curing times (15 and 540 days); iii) SEM characterization at different curing times (28 and 540 days); iv) the trend of the solution conductivity and pH during the leaching test; v) the characterization of the short-range ordering in terms of T−O−T bonds (where T is Al or Si) by 29 Si and 27 Al solid state magic-angle spinning nuclear magnetic resonance (ss MAS NMR) for geopolymers

  15. Chromium liquid waste inertization in an inorganic alkali activated matrix: Leaching and NMR multinuclear approach

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni, Chiara, E-mail: chiara.ponzoni@unimore.it [University of Modena and Reggio Emilia, Department of Engineering “Enzo Ferrari”, Modena (Italy); Lancellotti, Isabella; Barbieri, Luisa [University of Modena and Reggio Emilia, Department of Engineering “Enzo Ferrari”, Modena (Italy); Spinella, Alberto; Saladino, Maria Luisa [University of Palermo CGA-UniNetLab, Palermo (Italy); Martino, Delia Chillura [University of Palermo, Department STEBICEF, Palermo (Italy); Caponetti, Eugenio [University of Palermo CGA-UniNetLab, Palermo (Italy); University of Palermo, Department STEBICEF, Palermo (Italy); Armetta, Francesco [University of Palermo, Department STEBICEF, Palermo (Italy); Leonelli, Cristina [University of Modena and Reggio Emilia, Department of Engineering “Enzo Ferrari”, Modena (Italy)

    2015-04-09

    Highlights: • Inertization of chromium liquid waste in aluminosilicate matrix. • Water less inertization technique exploiting the waste water content. • Liquid waste inertization without drying step. • Long term stabilization study through leaching test. • SEM analysis and {sup 29}Si and {sup 27}Al MAS NMR in relation with long curing time. - Abstract: A class of inorganic binders, also known as geopolymers, can be obtained by alkali activation of aluminosilicate powders at room temperature. The process is affected by many parameters (curing time, curing temperature, relative humidity etc.) and leads to a resistant matrix usable for inertization of hazardous waste. In this study an industrial liquid waste containing a high amount of chromium (≈2.3 wt%) in the form of metalorganic salts is inertized into a metakaolin based geopolymer matrix. One of the innovative aspects is the exploitation of the water contained in the waste for the geopolymerization process. This avoided any drying treatment, a common step in the management of liquid hazardous waste. The evolution of the process - from the precursor dissolution to the final geopolymer matrix hardening - of different geopolymers containing a waste amount ranging from 3 to 20% wt and their capability to inertize chromium cations were studied by: i) the leaching tests, according to the EN 12,457 regulation, at different curing times (15, 28, 90 and 540 days) monitoring releases of chromium ions (Cr(III) and Cr(VI)) and the cations constituting the aluminosilicate matrix (Na, Si, Al); ii) the humidity variation for different curing times (15 and 540 days); iii) SEM characterization at different curing times (28 and 540 days); iv) the trend of the solution conductivity and pH during the leaching test; v) the characterization of the short-range ordering in terms of T−O−T bonds (where T is Al or Si) by {sup 29}Si and {sup 27}Al solid state magic-angle spinning nuclear magnetic resonance (ss MAS NMR) for

  16. Overview on the Multinational Collaborative Waste Storage and Disposal Solutions

    International Nuclear Information System (INIS)

    MARGEANU, C.A.

    2013-01-01

    The main drivers for a Safe, Secure and Global Energy future become clear and unequivocal: Security of supply for energy sources, Low-carbon electricity generation and Extended nuclear power assuring economic nuclear energy production, safe nuclear facilities and materials, safe and secure radioactive waste management and public acceptance. Responsible use of nuclear power requires that – in addition to safety, security and environmental protection associated with NPPs operation – credible solutions to be developed for dealing with the radioactive waste produced and especially for a responsible long term radioactive waste management. The paper deals with the existing multinational initiative in nuclear fuel cycle and the technical documents sustaining the multinational/regional disposal approach. Meantime, the paper far-reaching goal is to highlight on: What is offering the multinational waste storage and disposal solutions in terms of improved nuclear security ‽

  17. 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

  18. Determination of plutonium 241 in solutions of nuclear wastes

    International Nuclear Information System (INIS)

    Raymond, A.; Bilcot, J.B.; Poletiko, C.

    1990-09-01

    Determination of plutonium 241 in nuclear wastes is important because of long period and high energy of some daughter products. In this report are presented two quantitative analysis methods using both scintillation techniques: A complete method, in any case, by selective extraction of plutonium on an anionic resin allowing simultaneous determination of Pu 241 and the sum of other plutonium isotopes; a simplified method when alpha activity is higher than beta/gamma activity by liquid extraction with TTA. These methods are applied for analysis of 4 waste types: cement encapsulated wastes, bitumen encapsulated wastes, incineration ashes, leaching of encapsulated incineration ashes. In these 4 examples, Pu 241 activity is equal or higher than the sum of alpha plutonium isotope activity. Separation efficiency, measured from Pu 239 or with Pu 236 as tracer, is between 90 and 99% [fr

  19. Recovery of cyanide in gold leach waste solution by volatilization and absorption.

    Science.gov (United States)

    Gönen, N; Kabasakal, O S; Ozdil, G

    2004-09-10

    In this study, the effects of pH, time and temperature in regeneration of cyanide in the leaching waste solution of gold production from disseminated gold ore by cyanidation process were investigated and the optimum conditions, consumptions and cyanide recovery values were determined. The sample of waste solution containing 156 mg/l free CN- and 358 mg/l total CN-, that was obtained from Gümüşhane-Mastra/Turkey disseminated gold ores by cyanidation and carbon-in-pulp (CIP) process under laboratory conditions was used in the experiments. Acidification with H2SO4, volatilization of hydrogen cyanide (HCN) with air stripping and absorption of HCN in a basic solution stages were applied and under optimum conditions, 100% of free cyanide and 48% of complex cyanide and consequently 70% of the total cyanide in the liquid phase of gold leach effluent are recovered.

  20. Separation and recovery of ruthenium from radioactive liquid waste for specific medical applications - wealth from waste

    International Nuclear Information System (INIS)

    Pente, A.S.; Ramchandran, M.; Wawale, P.R.; Thorat, Vidya; Gireesan, Prema; Katarni, V.G.; Kumar, Amar; Kaushik, C.P.; Raj, Kanwar

    2010-01-01

    In recent past, 106 Ru has emerged as one of the promising β - emitting radionuclide used in brachytherapy for the treatment of choroidal melanoma and retinoblastoma due to its favorable nuclear decay characteristics. A plaque with low amount of 106 Ru activity of the order of 12 - 26 MBq (0.3 - 0.7 mCi ) is suitable for the above treatment and can be used for an adequate duration of 1-2 years due to suitable half-life (T 1/2 = 1.02 y). In order to undertake the preparation of 106 Ru plaque, an indigenous availability of this radionuclide with acceptable purity was explored from radioactive liquid waste having wide spectrum of fission products in line with wealth from waste strategy. Process methodology has been developed and standardized at Process Control Laboratory of Waste Immobilization Plant (WIP), Trombay for separation of 106 Ru from radioactive liquid waste for intended medical application. (author)

  1. Assessment of industrial liquid waste management in Omdurman Industrial Area

    International Nuclear Information System (INIS)

    Elnasri, R. A. A.

    2003-04-01

    This study was conducted mainly to investigate the effects of industrial liquid waste on the environment in the Omdurman area. Various types of industries are found around Omdurman. According to the ISC the major industries are divided into eight major sub-sectors, each sub-sector is divided into types of industries. Special consideration was given to the liquid waste because of its effects. In addition to the available data, personal observation supported by photographs, laboratory analyses were carried on the industrial effluents. The investigated parameters in the analysis were, BOD, COD, O and G, Cr, TDS, TSS, pH, temp and conductivity. Interviews were conducted with waste handling workers in the industries, in order to assess the effects of industrial pollution. The results obtained showed that pollutants produced by all the factories were found to exceed the accepted levels of the industrial pollution control. The effluents disposed of in the sites allotted by municipal authorities have adverse effects on the surrounding environment and public health and amenities. Accordingly the study recommends that the waste water must be pretreated before being disposed of in site allotted by municipal authorities. Develop an appropriate system for industrial waste proper management. The study established the need to construct a sewage system in the area in order to minimize the pollutants from effluents. (Author)

  2. Assessment of industrial liquid waste management in Omdurman Industrial Area

    Energy Technology Data Exchange (ETDEWEB)

    Elnasri, R A. A. [Institute of Environmental Studies, University of Khartoum, Khartoum (Sudan)

    2003-04-15

    This study was conducted mainly to investigate the effects of industrial liquid waste on the environment in the Omdurman area. Various types of industries are found around Omdurman. According to the ISC the major industries are divided into eight major sub-sectors, each sub-sector is divided into types of industries. Special consideration was given to the liquid waste because of its effects. In addition to the available data, personal observation supported by photographs, laboratory analyses were carried on the industrial effluents. The investigated parameters in the analysis were, BOD, COD, O and G, Cr, TDS, TSS, pH, temp and conductivity. Interviews were conducted with waste handling workers in the industries, in order to assess the effects of industrial pollution. The results obtained showed that pollutants produced by all the factories were found to exceed the accepted levels of the industrial pollution control. The effluents disposed of in the sites allotted by municipal authorities have adverse effects on the surrounding environment and public health and amenities. Accordingly the study recommends that the waste water must be pretreated before being disposed of in site allotted by municipal authorities. Develop an appropriate system for industrial waste proper management. The study established the need to construct a sewage system in the area in order to minimize the pollutants from effluents. (Author)

  3. Turbulence effects on volatilization rates of liquids and solutes.

    Science.gov (United States)

    Lee, Jiunn-Fwu; Chao, Huan-Ping; Chiou, Cary T; Manes, Milton

    2004-08-15

    Volatilization rates of neat liquids (benzene, toluene, fluorobenzene, bromobenzene, ethylbenzene, m-xylene, o-xylene, o-dichlorobenzene, and 1-methylnaphthalene) and of solutes (phenol, m-cresol, benzene, toluene, ethylbenzene, o-xylene, and ethylene dibromide) from dilute water solutions have been measured in the laboratory over a wide range of air speeds and water-stirring rates. The overall transfer coefficients (K(L)) for individual solutes are independent of whether they are in single- or multi-solute solutions. The gas-film transfer coefficients (kG) for solutes in the two-film model, which have hitherto been estimated by extrapolation from reference coefficients, can now be determined directly from the volatilization rates of neat liquids through a new algorithm. The associated liquid-film transfer coefficients (kL) can then be obtained from measured K(L) and kG values and solute Henry law constants (H). This approach provides a novel means for checking the precision of any kL and kG estimation methods for ultimate prediction of K(L). The improved kG estimation enables accurate K(L) predictions for low-volatility (i.e., low-H) solutes where K(L) and kGH are essentially equal. In addition, the prediction of K(L) values for high-volatility (i.e., high-H) solutes, where K(L) approximately equal to kL, is also improved by using appropriate reference kL values.

  4. Removal of radioactive materials from waste solutions via magnetic ferrites

    International Nuclear Information System (INIS)

    Boyd, T.E.; Kochen, R.L.; Price, M.Y.

    1982-01-01

    Ferrite waste treatment was found to be effective in removing actinides from simulated Rocky Flats process waste solutions. With a one-stage ferrite treatment, plutonium concentrations were consistently reduced from 10 -4 g/l to less than 10 -8 g/l, and americium concentrations were lowered from 10 -7 g/l to below 10 -10 g/l. In addition, siginficantly less solid was produced as compared with the flocculant precipitation technique now employed at Rocky Flats. Aging of ferrite solids and elevated beryllium and phosphate concentrations were identified as interferences in the ferrite treatment of process waste, but neither appeeared serious enough to prevent implementation in plant operations

  5. Methods for removing transuranic elements from waste solutions

    International Nuclear Information System (INIS)

    Slater, S.A.; Chamberlain, D.B.; Connor, C.; Sedlet, J.; Srinivasan, B.; Vandegrift, G.F.

    1994-11-01

    This report outlines a treatment scheme for separating and concentrating the transuranic (TRU) elements present in aqueous waste solutions stored at Argonne National Laboratory (ANL). The treatment method selected is carrier precipitation. Potential carriers will be evaluated in future laboratory work, beginning with ferric hydroxide and magnetite. The process will result in a supernatant with alpha activity low enough that it can be treated in the existing evaporator/concentrator at ANL. The separated TRU waste will be packaged for shipment to the Waste Isolation Pilot Plant

  6. Productive Liquid Fertilizer from Liquid Waste Tempe Industry as Revealed by Various EM4 Concentration

    Science.gov (United States)

    Hartini, S.; Letsoin, F.; Kristijanto, A. I.

    2018-04-01

    Recently, using of productive liquid fertilizer assumed as a proper and practical fertilizer for plant productivity purposes. Various ways of enrichment of liquid fertilizer were done to achieve certain quality. The purpose of this research was to determine the proper additional formulation in the process of making productive liquid fertilizer based on the various concentration of EM4 as well as comparated the result with SNI. Liquid tempe waste were collected from some tempe industries at Sidorejo Kidul village, Tingkir district, Salatiga. The concentration of EM4 which were added to the tempe wastewater are 0%; 0.20%; 0.40%; 0.60%; 0.80%; 1.00% respectively. The pH, temperature, C total, N total, C/N ratio, and PO4 3- were measured. Data was analyzed by using Randomize Completely Block Design (RCBD) with 6 treatments and 4 replications. Comparison between the average, the Honestly Significance Deference (HSD) 5% was used. The results showed that the addition of EM4 indicated there were a significant progress. Moreover, the most effective formula to increase the quality of productive liquid fertilizer from liquid waste tempe was found in addition of 1.00% EM4 with the gained analysis value for the C total, N total, C/N ratio, and degree of PO4 3- as follows : 4.395 ± 1.034%; 1.470 ± 0.081%; 3.01 ± 0.756; 685.28 ± 70.44 ppm . Associated with the need fulfillment of SNI hence can be concluded that result of Productive Liquid Fertilizer (PLF) from liquid waste tempe successfully fulfill SNI of liquid fertilizer for pH parameter and total N, only.

  7. Treatment systems for liquid wastes generated in chemical analysis laboratories

    International Nuclear Information System (INIS)

    Linda Berrio; Oscar Beltran; Edison Agudelo; Santiago Cardona

    2012-01-01

    Nowadays, handling of liquid wastes from chemical analysis laboratories is posing problems to different public and private organizations because of its requirements of an integrated management. This article reviews various treatment technologies and its removal efficiencies in order to establish criteria for selecting the system and the appropriate variables to achieve research objectives as well as environmental sustainability. Review begins with a description of the problem and continues with the study of treatments for laboratory wastes. These technologies are segregated into physicochemical and biological treatments that comprise a variety of processes, some of which are considered in this review.

  8. Screening of Acetic Acid Bacteria from Pineapple Waste for Bacterial Cellulose Production using Sago Liquid Waste

    Directory of Open Access Journals (Sweden)

    Nur Arfa Yanti

    2017-12-01

    Full Text Available Bacterial cellulose is a biopolymer produced by fermentation process with the help of bacteria. It has numerous applications in industrial sector with its characteristic as a biodegradable and nontoxic compound in nature. The potential application of BC is limited by its production costs, because BC is produced from expensive culture media. The use of cheap carbon and nutrient sources such as sago liquid waste is an interesting strategy to overcome this limitation. The objective of this study was to obtain the AAB strain that capable to produce bacterial cellulose from sago liquid waste. Isolation of AAB strains was conducted using CARR media and the screening of BC production was performed on Hestrin-Schramm (HS media with glucose as a carbon source. The strains of AAB then were evaluated for their cellulose-producing capability using sago liquid waste as a substrate. Thirteen strains of AAB producing BC were isolated from pineapple waste (pineapple core and peel and seven of them were capable to produce BC using sago liquid waste substrate. One of the AAB strains produced a relatively high BC, i.e. isolate LKN6. The result of morphological and biochemical test was proven that the bacteria was Acetobacter xylinum. The result of this study showed that A. xylinum LKN6 can produce a high yield of BC, therefore this strain is potentially useful for its utilization as a starter in bacterial cellulose production. 

  9. Quantity assessment of waste in the dismantlement of liquid waste treatment plant and its actual state

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Mitsuhashi, Ishi; Matsumoto, Tetsuo; Morishima, Kayoko; Tanzawa, Tomio

    2016-01-01

    From the progress of decommissioning project work of Tokyo City University Atomic Energy Research Institute, this paper reports the comparison between the actual amount of the waste generated during dismantlement work at liquid waste treatment facilities and the assessment quantity before starting the dismantlement. The quantity assessment was made on the basis of the installation license application, design specifications, drawings, records, history of use, site investigation results, etc. Since this quantity assessment did not take into account the dismantling contents of reservoir concrete, the assessed quantity of non-radioactive waste (NR waste) did not match the sum of actual NR waste. However, if an actually generated quantity of concrete of radioactive waste was added to the quantity assessment as NR waste, the quantity of actually generated NR waste and that of assessed NR waste were nearly consistent, which verified the validity of this assessment. This method is considered to be able to be utilized in the future quantity assessment of decommissioning work and the like. On the other hand, it was found that the number of drums that were actually stored tended to increase more than the estimated number of drum conversion. In old buildings, it is necessary to take into account the generation of waste other than radioactive materials in the quantity assessment stage and dismantlement stage. (A.O.)

  10. Behavior of supercooled aqueous solutions stemming from hidden liquid-liquid transition in water.

    Science.gov (United States)

    Biddle, John W; Holten, Vincent; Anisimov, Mikhail A

    2014-08-21

    A popular hypothesis that explains the anomalies of supercooled water is the existence of a metastable liquid-liquid transition hidden below the line of homogeneous nucleation. If this transition exists and if it is terminated by a critical point, the addition of a solute should generate a line of liquid-liquid critical points emanating from the critical point of pure metastable water. We have analyzed thermodynamic consequences of this scenario. In particular, we consider the behavior of two systems, H2O-NaCl and H2O-glycerol. We find the behavior of the heat capacity in supercooled aqueous solutions of NaCl, as reported by Archer and Carter [J. Phys. Chem. B 104, 8563 (2000)], to be consistent with the presence of the metastable liquid-liquid transition. We elucidate the non-conserved nature of the order parameter (extent of "reaction" between two alternative structures of water) and the consequences of its coupling with conserved properties (density and concentration). We also show how the shape of the critical line in a solution controls the difference in concentration of the coexisting liquid phases.

  11. Liquid and Gaseous Waste Operations Department annual operating report CY 1996

    International Nuclear Information System (INIS)

    Maddox, J.J.; Scott, C.B.

    1997-03-01

    This annual report summarizes operating activities dealing with the process waste system, the liquid low-level waste system, and the gaseous waste system. It also describes upgrade activities dealing with the process and liquid low-level waste systems, the cathodic protection system, a stack ventilation system, and configuration control. Maintenance activities are described dealing with nonradiological wastewater treatment plant, process waste treatment plant and collection system, liquid low-level waste system, and gaseous waste system. Miscellaneous activities include training, audits/reviews/tours, and environmental restoration support

  12. Process for producing zeolite adsorbent and process for treating radioactive liquid waste with the zeolite adsorbent

    International Nuclear Information System (INIS)

    Motojima, K.; Kawamura, F.

    1984-01-01

    Zeolite is contacted with an aqueous solution containing at least one of copper, nickel, cobalt, manganese and zinc salts, preferably copper and nickel salts, particularly preferably copper salt, in such a form as sulfate, nitrate, or chloride, thereby adsorbing the metal on the zeolite in its pores by ion exchange, then the zeolite is treated with a water-soluble ferrocyanide compound, for example, potassium ferrocyanide, thereby forming metal ferrocyanide on the zeolite in its pores. Then, the zeolite is subjected to ageing treatment, thereby producing a zeolite adsorbent impregnated with metal ferrocyanide in the pores of zeolite. The adsorbent can selectively recover cesium with a high percent cesium removal from a radioactive liquid waste containing at least radioactive cesium, for example, a radioactive liquid waste containing cesium and such coexisting ions as sodium, magnesium, calcium and carbonate ions at the same time at a high concentration. The zeolite adsorbent has a stable adsorbability for a prolonged time

  13. 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

  14. Molten salt hazardous waste disposal process utilizing gas/liquid contact for salt recovery

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.

    1984-01-01

    The products of a molten salt combustion of hazardous wastes are converted into a cooled gas, which can be filtered to remove hazardous particulate material, and a dry flowable mixture of salts, which can be recycled for use in the molten salt combustion, by means of gas/liquid contact between the gaseous products of combustion of the hazardous waste and a solution produced by quenching the spent melt from such molten salt combustion. The process results in maximizing the proportion of useful materials recovered from the molten salt combustion and minimizing the volume of material which must be discarded. In a preferred embodiment a spray dryer treatment is used to achieve the desired gas/liquid contact

  15. The selective removal of 90Sr and 137Cs from liquid low-level waste at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Bostick, D.T.; Arnold, W.D.; Burgess, M.W.; Taylor, P.A.; Kent, T.E.

    1995-01-01

    Methods are being developed for the selective removal of the two principal radioactive contaminants, 90 Sr and 137 Cs, from liquid low-level waste generated and/or stored at Oak Ridge National Laboratory. These methods are to be used in a future centralized treatment facility at ORNL. Removal of 90 Sr in the proposed treatment flashed is based on coprecipitation from strongly alkaline waste by adding stable strontium to the waste solution. Ferric sulfate, added with the stable strontium, improves the 90 Sr removal and aids in the flocculation of the strontium carbonate (SrCO 3 ) precipitate. After separation of the solids, the resultant supernate is adjusted to pH 8 for the cesium removal treatment. Upon pH adjustment, aluminum originally present in the untreated alkaline waste precipitates and sorbs an additional amount of 90 Sr. Cesium is removed from the neutralized waste by two sequential treatments with potassium cobalt hexacyanoferrate (KCCF) slurry formed by the addition of potassium ferrocyanide (K 4 Fe(CN) 6 ) and cobalt nitrate (Co(NO 3 ) 2 ) solutions. The cumulative decontamination factors (DFs) for 90 Sr and 137 Cs in benchscale studies are 4900 and 1 x 10 6 , respectively, if high speed centrifugation is used for the liquid/solid separations. Efforts are now underway to evaluate process-scale techniques to perform the liquid/solid separations required for removal of SrCO 3 and 137 Cs-bearing hexacyanoferrate solids from the treated waste solution

  16. Application of membrane technologies for liquid radioactive waste processing

    International Nuclear Information System (INIS)

    2004-01-01

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

  17. 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.)

  18. Proceedings of the international seminar on chemistry and process engineering for high-level liquid waste solidification

    International Nuclear Information System (INIS)

    Odoj, R.; Merz, E.

    1981-06-01

    The proceedings record a very distinct phase of the chemistry and process engineering for high-level liquid waste solidification in the past years. The main purpose is to provide solutions which guarantee sufficient safe and economically acceptable measure causing no adverse consequence to man and his environment. (DG)

  19. Diglycolamide-functionalized task specific ionic liquids for nuclear waste remediation: extraction, luminescence, theoretical and EPR investigations

    NARCIS (Netherlands)

    Sengupta, A; Mohapatra, P.K.; Kadam, R.M.; Manna, D.; Ghanty, T.K.; Iqbal, M.; Huskens, Jurriaan; Verboom, Willem

    2014-01-01

    A 3.6 × 10−2 M solution of a diglycolamide-functionalized task specific ionic liquid (DGA-TSIL) in [C4mim][NTf2] was used for the extraction of actinides (mainly Am) and other elements present in high level nuclear waste. The extraction of Eu3+ was relatively higher than that of Am3+ conforming to

  20. The study of sorption of cesium radionuclides by 'T-55' ferrocyanide sorbent from various types of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Semenischev, V.S.; Voronina, A.V.; Bykov, A.A.

    2013-01-01

    The sorption of caesium by T-55 sorbent from different types of liquid radioactive wastes is studied. It is shown that the sorbent can be used for extraction of cesium from high level acidic and saline solutions and also for decontamination of caesium contaminated waters containing surfactants and EDTA. (author)

  1. 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)

  2. Analysis Of Liquid Waste Management At Dr. Mohammad Hoesin Palembang's Hospital

    OpenAIRE

    Hartini, Resi; Hasyim, Hamzah; Ainy, Asmaripa

    2011-01-01

    Background : The hospital is an institution that service activities of preventive, curative, rehabilitative and promotive health. These activities produce solid, liquid, and gas waste. Liquid waste can cause diseases and environment pollution so need special waste management. Dr. Mohammad Hoesin Palembang's Hospital producea lot of liquid waste. Method : This study is a descriptive research with qualitative approach. Sources of information consist four informants. The research are using dept...

  3. Pharmaceutical Perspective on Opalescence and Liquid-Liquid Phase Separation in Protein Solutions.

    Science.gov (United States)

    Raut, Ashlesha S; Kalonia, Devendra S

    2016-05-02

    Opalescence in protein solutions reduces aesthetic appeal of a formulation and can be an indicator of the presence of aggregates or precursor to phase separation in solution signifying reduced product stability. Liquid-liquid phase separation of a protein solution into a protein-rich and a protein-poor phase has been well-documented for globular proteins and recently observed for monoclonal antibody solutions, resulting in physical instability of the formulation. The present review discusses opalescence and liquid-liquid phase separation (LLPS) for therapeutic protein formulations. A brief discussion on theoretical concepts based on thermodynamics, kinetics, and light scattering is presented. This review also discusses theoretical concepts behind intense light scattering in the vicinity of the critical point termed as "critical opalescence". Both opalescence and LLPS are affected by the formulation factors including pH, ionic strength, protein concentration, temperature, and excipients. Literature reports for the effect of these formulation factors on attractive protein-protein interactions in solution as assessed by the second virial coefficient (B2) and the cloud-point temperature (Tcloud) measurements are also presented. The review also highlights pharmaceutical implications of LLPS in protein solutions.

  4. EB detoxification of liquid hazardous wastes

    International Nuclear Information System (INIS)

    Tata, A.; Giuliani, S.

    1996-07-01

    In the work, an engineering approach to technical solutions, considering accelerated electron beams as radiation source, is carried out, in order to allow and evaluate an effective recovery of drinking water from highly chemically polluted groundwaters. In connection with different engineering technical and economic parameters (suitable doses, EB-machine type, plant features, etc.) and with reference to 1-100 ton/hr contaminated stream flowrate range (2-50 kGy as considered absorbed dose range), a specifically developed computer code was run. Analysis results, based on investment and functioning cost figures evaluated with reference to industrial plant management scenarios, are treatment unit costs showing a noticeable economic attractiveness of radiation EB-technologies in the field of considered applications

  5. Properties of gases, liquids, and solutions principles and methods

    CERN Document Server

    Mason, Warren P

    2013-01-01

    Physical Acoustics: Principles and Methods, Volume ll-Part A: Properties of Gases, Liquids, and Solutions ponders on high frequency sound waves in gases, liquids, and solids that have been proven as effective tools in examining the molecular, domain wall, and other types of motions. The selection first offers information on the transmission of sound waves in gases at very low pressures and the phenomenological theory of the relaxation phenomena in gases. Topics include free molecule propagation, phenomenological thermodynamics of irreversible processes, and simultaneous multiple relaxation pro

  6. NMR and molecular dynamics of small solutes in liquid crystals

    International Nuclear Information System (INIS)

    Luyten, P.R.

    1984-01-01

    NMR relaxation measurements, using a wide variety of modern pulse techniques, can yield valuable information about molecular motions. In this thesis the applicability of theories developed to describe spin relaxation phenomena in partially ordered media is studied for small solutes in liquid crystals. 1 H, 2 H, 13 C and 14 N relaxation measurements are interpreted by means of a model, in which fast anisotropic re-orientational motion in an orienting potential combined with contributions from cooperative fluctuations in the surrounding liquid crystal molecules, induce the observed frequency dependent relaxation behavior. (orig.)

  7. 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.)

  8. Photoionization of Sodium Salt Solutions in a Liquid Jet

    Energy Technology Data Exchange (ETDEWEB)

    Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

    2008-06-05

    A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces.

  9. Photoionization of Sodium Salt Solutions in a Liquid Jet

    International Nuclear Information System (INIS)

    Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

    2008-01-01

    A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces

  10. Seismic evaluation of existing liquid low level waste system at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Hammond, C.R.; Holmes, R.M.; Kincaid, J.H.; Singhal, M.K.; Stockdale, B.I.; Walls, J.C.; Webb, D.S.

    1993-01-01

    The existing liquid low level waste (LLLW) system at the Oak Ridge National Laboratory is used to collect, neutralize, concentrate, and store the radioactive and toxic waste from various sources at the Laboratory. The waste solutions are discharged from source facilities to individual collection tanks, transferred by underground piping to an evaporator facility for concentration, and pumped through the underground piping to storage in underground tanks. The existing LLLW system was installed in the 1950s with several system additions up to the present. The worst-case accident postulated is an earthquake of sufficient magnitude to rupture the tanks and/or piping so as to damage the containment integrity to the surrounding soil and environment. The objective of an analysis of the system is to provide a level of confidence in the seismic resistance of the LLLW system to withstand the postulated earthquake

  11. 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)

  12. Solutions for energy recovery of animal waste from leather industry

    International Nuclear Information System (INIS)

    Lazaroiu, Gheorghe; Pană, Constantin; Mihaescu, Lucian; Cernat, Alexandru; Negurescu, Niculae; Mocanu, Raluca; Negreanu, Gabriel

    2017-01-01

    Highlights: • Animal fats in blend with diesel fuel for energy valorification through combustion. • Animal waste from tanneries as fuel and for biogas production. • Experimental tests using animal fats as fuel for diesel engines. • Experimental tests modifying the characteristic parameters. - Abstract: Secondary products from food and leather industries are regarded as animal wastes. Conversion of these animal wastes into fuels represents an energy recovery solution not only because of their good combustion properties, but also from the viewpoint of supply stability. A tannery factory usually processes 60–70 t/month of crude leathers, resulting in 12–15 t/month of waste. Fats, which can be used as the input fuel for diesel engines (in crude state or as biodiesel), represent 10% of this animal waste, while the rest are proteins that can be used to generate biogas through anaerobic digestion. Herein, we analyse two approaches to the use of animal waste from tanneries: as fuel for diesel engines and for biogas generation for heat production. Diesel fuelling and fuelling by animal wastes are compared in terms of the engine performance and pollutant emissions. The effects of animal waste usage on the pollutant emissions level, exhaust gas temperature, indicated mean effective pressure, maximum pressure, and engine efficiency are analysed. The energy recovery technologies for animal waste, which are analysed in this work, can be easily implemented and can simultaneously solve the problem posed by animal wastes by using them as an alternative to fossil fuels. Animal fats can be considered an excellent alternative fuel for diesel engines without major constructive modifications.

  13. Dissolution of agro-waste in ionic liquids

    International Nuclear Information System (INIS)

    Lee, Kiat Moon; Ngoh, Gek Cheng; Chua, Adeline Seak May

    2010-01-01

    Full text: There are abundant of agro-wastes being produced in Malaysia. One of the largely produced agro wastes is the sago hampas. It is known as a strong environmental pollutant due to its cellulosic fibrous material. However, the presence of the starch, cellulose and hemicelluloses in the hampas can be converted into valuable products such as reducing sugars. Hence, this study was performed to investigate the ability of ionic liquids in hydrolysing the ligno celluloses biomass into reducing sugars. Three types of ionic liquids were used, 1-butyl-3-methylimidazolium chloride (BMIM Cl), 1-ethyl-3- methylimidazolium acetate (EMIM Ac) and 1-ethyl-3-methylimidazolium diethyl phosphate (EMIM DEP). The reaction was performed by heating the reaction mixture of sago hampas and ionic liquids at 100 degree Celsius. The concentrations of reducing sugars in the hydrolysates were determined by DNS method. Maximum concentration of reducing sugars were 0.424, 0.299, 0.260 mg/ml for BmimCl, EmimAc and EmimDEP respectively. These concluded that the selected ionic liquids were inefficient in hydrolysing the sago hampas to reducing sugars. (author)

  14. Idaho Nuclear Technology and Engineering Center Newly Generated Liquid Waste Demonstration Project Feasibility Study

    International Nuclear Information System (INIS)

    Herbst, A.K.

    2000-01-01

    A research, development, and demonstration project for the grouting of newly generated liquid waste (NGLW) at the Idaho Nuclear Technology and Engineering Center is considered feasible. NGLW is expected from process equipment waste, decontamination waste, analytical laboratory waste, fuel storage basin waste water, and high-level liquid waste evaporator condensate. The potential grouted waste would be classed as mixed low-level waste, stabilized and immobilized to meet RCRA LDR disposal in a grouting process in the CPP-604 facility, and then transported to the state

  15. Analysis of 99Tc in the radioactive liquid waste after extraction into suitable solvent

    International Nuclear Information System (INIS)

    Sonar, N.L.; Vaishali De; Pardeshi, V.; Raghvendra, Y.; Valsala, T.P.; Sonavane, M.S.; Kulkarni, Y.; Raj Kanwar

    2012-01-01

    99 Tc is one of the long lived fission product with high fission yield. >From radioactive waste management point of view it is very much essential to evaluate the concentration of technetium in the radioactive liquid waste in order to finalise the treatment process to extract/isolate it from the stream which is discharged to the environment. For the estimation of 99 Tc in the radioactive liquid waste stream, extraction of the stable complex of technetium-tetraphenyl arsonium chloride (TPAC) into chloroform followed by beta counting was studied. Various parameters like pH, time of equilibration, concentration of TPAC in chloroform, use of other solvent for extraction as well as interference of various other radionuclides present in the waste were also studied. The radioactive liquid waste being handled in plant contains high concentrations of salts in the form of sodium nitrate. Hence effect of salt concentration on the percentage extraction was also evaluated. The extraction behavior does not dependent on change in the pH of the solution. Almost 99.5% extraction was observed in the pH range of 1-13.0. High concentration of salt is affecting the extraction. However, this can be taken care by diluting the radioactive waste. It takes almost 90 min time for maximum extraction. Presence of radionuclides like 137 Cs, 90 Sr are not interfering the extraction of 99 Tc. However, 106 Ru is getting slightly extracted along with 99 Tc. The error due to 106 Ru can be eliminated by taking gamma spectrum and deducting the activity from the total beta activity to get 99 Tc activity. Nitrobenzene can be used for extraction of Tc-TPAC complex in place of chloroform. (author)

  16. Radioactive Waste...The Problem and Some Possible Solutions

    Science.gov (United States)

    Olivier, Jean-Pierre

    1977-01-01

    Nuclear safety is a highly technical and controversial subject that has caused much heated debate and political concern. This article examines the problems involved in managing radioactive wastes and the techniques now used. Potential solutions are suggested and the need for international cooperation is stressed. (Author/MA)

  17. Food waste in Central Europe – challenges and solutions

    Directory of Open Access Journals (Sweden)

    den Boer Jan

    2017-01-01

    Full Text Available Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland, founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  18. Food waste in Central Europe - challenges and solutions

    Science.gov (United States)

    den Boer, Jan; Kobel, Przemysław; Dyjakon, Arkadiusz; Urbańska, Klaudia; Obersteiner, Gudrun; Hrad, Marlies; Schmied, Elisabeth; den Boer, Emilia

    2017-11-01

    Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe) was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland), founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  19. The application of experimental design methodology for the investigation of liquid radioactive waste treatment

    Directory of Open Access Journals (Sweden)

    Šljivić-Ivanović Marija Z.

    2017-01-01

    Full Text Available The sorption properties of waste facade, brick, and asphalt sample towards Sr(II, Co(II, and Ni(II ions from single and multicomponent solutions were investigated. The highest sorption capacity was found for Ni(II ions, while the most effective sorbent was facade. Simplex Centroid Mixture Design was used in order to investigate the sorption processes of ions from solutions with different composition as well as the competition between the cations. Based on the statistical analysis results, the equations for data modeling were proposed. According to the observations, the investigated solid matrices can be effectively used for the liquid radioactive waste treatment. Furthermore, the applied methodology turned out to be an easy and operational way for the investigations of multicomponent sorption processes. [Project of the Serbian Ministry of Education, Science and Technological Development, Grant no. III 43009 and Grant no. OI 171007

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. Selective extraction and recovery of rare earth metals from phosphor powders in waste fluorescent lamps using an ionic liquid system

    International Nuclear Information System (INIS)

    Yang, Fan; Kubota, Fukiko; Baba, Yuzo; Kamiya, Noriho; Goto, Masahiro

    2013-01-01

    Highlights: • Recycling of rare earth metals from fluorescent lamps was conducted by ionic liquid-mediated extraction. • Acid leaching from a waste phosphor powder was carried out using sulfuric and nitric acids. • An ionic liquid was used as extracting solvent for the rare earth metals. • Selective extraction of rare earth metals from leach solutions was attained. •The extracting ionic liquid phase was recyclable in the recovery process. -- Abstract: The recycling of rare earth metals from phosphor powders in waste fluorescent lamps by solvent extraction using ionic liquids was studied. Acid leaching of rare earth metals from the waste phosphor powder was examined first. Yttrium (Y) and europium (Eu) dissolved readily in the acid solution; however, the leaching of other rare earth metals required substantial energy input. Ionization of target rare earth metals from the waste phosphor powders into the leach solution was critical for their successful recovery. As a high temperature was required for the complete leaching of all rare earth metals, ionic liquids, for which vapor pressure is negligible, were used as an alternative extracting phase to the conventional organic diluent. An extractant, N, N-dioctyldiglycol amic acid (DODGAA), which was recently developed, showed a high affinity for rare earth metal ions in liquid–liquid extraction although a conventional commercial phosphonic extractant did not. An effective recovery of the rare earth metals, Y, Eu, La and Ce, from the metal impurities, Fe, Al and Zn, was achieved from the acidic leach solution of phosphor powders using an ionic liquid containing DODGAA as novel extractant system

  2. Selective extraction and recovery of rare earth metals from phosphor powders in waste fluorescent lamps using an ionic liquid system

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Fan; Kubota, Fukiko; Baba, Yuzo [Department of Applied Chemistry, Graduate School of Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Kamiya, Noriho [Department of Applied Chemistry, Graduate School of Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Center for Future Chemistry, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Goto, Masahiro, E-mail: m-goto@mail.cstm.kyushu-u.ac.jp [Department of Applied Chemistry, Graduate School of Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Center for Future Chemistry, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan)

    2013-06-15

    Highlights: • Recycling of rare earth metals from fluorescent lamps was conducted by ionic liquid-mediated extraction. • Acid leaching from a waste phosphor powder was carried out using sulfuric and nitric acids. • An ionic liquid was used as extracting solvent for the rare earth metals. • Selective extraction of rare earth metals from leach solutions was attained. •The extracting ionic liquid phase was recyclable in the recovery process. -- Abstract: The recycling of rare earth metals from phosphor powders in waste fluorescent lamps by solvent extraction using ionic liquids was studied. Acid leaching of rare earth metals from the waste phosphor powder was examined first. Yttrium (Y) and europium (Eu) dissolved readily in the acid solution; however, the leaching of other rare earth metals required substantial energy input. Ionization of target rare earth metals from the waste phosphor powders into the leach solution was critical for their successful recovery. As a high temperature was required for the complete leaching of all rare earth metals, ionic liquids, for which vapor pressure is negligible, were used as an alternative extracting phase to the conventional organic diluent. An extractant, N, N-dioctyldiglycol amic acid (DODGAA), which was recently developed, showed a high affinity for rare earth metal ions in liquid–liquid extraction although a conventional commercial phosphonic extractant did not. An effective recovery of the rare earth metals, Y, Eu, La and Ce, from the metal impurities, Fe, Al and Zn, was achieved from the acidic leach solution of phosphor powders using an ionic liquid containing DODGAA as novel extractant system.

  3. 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.

  4. Potential radiation damage: Storage tanks for liquid radioactive waste

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1992-01-01

    High level waste at SRS is stored in carbon steel tanks constructed during the period 1951 to 1981. This waste contains radionuclides that decay by alpha, beta, or gamma emission or are spontaneous neutronsources. Thus, a low intensity radiation field is generated that is capable of causing displacement damage to the carbon steel. The potential for degradation of mechanical properties was evaluated by comparing the estimated displacement damage with published data relating changes in Charpy V-notch (CVN) impact energy to neutron exposure. Experimental radiation data was available for three of the four grades of carbonsteel from which the tanks were constructed and is applicable to all four steels. Estimates of displacement damage arising from gamma and neutron radiation have been made based on the radionuclide contents for high level waste that are cited in the Safety Analysis Report (SAR) for the Liquid Waste Handling Facilities in the 200-Area. Alpha and beta emissions do not penetrate carbon steel to a sufficient depth to affect the bulk properties of the tank walls but may aggravate corrosion processes. The damage estimates take into account the source of the waste (F- or H-Area), the several types of tank service, and assume wateras an attenuating medium. Estimates of displacement damage are conservative because they are based on the highest levels of radionuclide contents reported in the SAR and continuous replenishment of the radionuclides

  5. The Recovery of Zinc Heavy Metal from Industrial Liquid Waste

    International Nuclear Information System (INIS)

    Panggabean, Sahat M.

    2000-01-01

    It had been studied the recovery of zinc heavy metal from liquid waste of electroplating industry located at East Jakarta. The aim of this study was to minimize the waste arisen from industrial activities by taking out zinc metal in order to reused on-site. The method of recovery was two steps precipitation using NaOH reagent and pH variation. The first step of precipitation at pH optimum around 6 yielded iron metal. The second step at pH optimum around 10 yielded zinc metal. The zinc metal was taken out assessed to the possibility of reused at that fabric. By applying its, it will yield the volume reduction of sludge waste about 36.1% or 53.2% of zinc metal containing in the waste. It means the cost of waste treatment will be lower. Beside its, the effluent arisen from the method had fulfill the maximum limit and it allowed to release to the environment. (author)

  6. Disposal of by-products in olive oil industry: waste-to-energy solutions

    International Nuclear Information System (INIS)

    Caputo, Antonio C.; Scacchia, Federica; Pelagagge, Pacifico M.

    2003-01-01

    Olive oil production industry is characterized by relevant amounts of liquid and solid by-products [olive mill wastewater (OMW) and olive husk (OH)], and by economical, technical and organizational constraints that make difficult the adoption of environmentally sustainable waste disposal approaches. In this context, waste treatment technologies aimed at energy recovery represent an interesting alternative. In the paper, a technical and economical analysis of thermal disposal plant solutions with energy recovery has been carried out. The considered plants enable the combined treatment of OMW and OH which, although penalizes the energy recovery, proves to be feasible and profitable in a future legislative scenario when stricter limitation on OMW disposal will force oil producers to bear high disposal costs. Results are compared by using economic performance measures, including revenues from produced energy and avoided disposal costs. A sensitivity and risk analysis is also performed in order to assess the economic profitability of the proposed solutions

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

    International Nuclear Information System (INIS)

    Alexandre, D.

    1986-05-01

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

  8. Liquid and Gaseous Waste Operations Department annual operating report CY 1994

    International Nuclear Information System (INIS)

    Maddox, J.J.; Scott, C.B.

    1995-03-01

    This report presents details about the operation of the liquid and gaseous waste department of Oak Ridge National Laboratory for the calendar year 1994. Topics discussed include; process waste system, upgrade activities, low-level liquid radioactive waste solidification project, maintenance activities, and other activities such as training, audits, and tours

  9. Fabrication of Greener Membranes from Ionic Liquid Solutions

    KAUST Repository

    Kim, DooLi

    2017-06-01

    Membrane technology plays a crucial role in different separation processes such as biotechnology, pharmaceutical, and food industries, drinking water supply, and wastewater treatment. However, there is a growing concern that solvents commonly used for membrane fabrication, such as dimethylformamide (DMF), dimethylacetamide (DMAc), and 1-methyl-2-pyrrolidone (NMP), are toxic to the environment and human health. To explore the possibility of substituting these toxic solvents by less toxic or safer solvents, polymers commonly used for membrane fabrication, such as polyacrylonitrile (PAN), cellulose acetate (CA), polyethersulfone (PES), and poly(ether imide sulfone) (EXTEMTM), were dissolved in ionic liquids. Flat sheet and hollow fiber membranes were then fabricated. The thermodynamics of the polymer solutions, the kinetics of phase inversion and other factors, which resulted in significant differences in the membrane structure, compared to those of membranes fabricated from more toxic solvents, were investigated. Higher water permeance with smaller pores, unique and uniform morphologies, and narrower pore size distribution, were observed in the ionic liquid-based membranes. Furthermore, comparable performance on separation of peptides and proteins with various molecular weights was achieved with the membranes fabricated from ionic liquid solutions. In summary, we propose less hazardous polymer solutions to the environment, which can be used for the membrane fabrication with better performance and more regular morphology.

  10. Determination of Na+ and K+ ions in the high-level liquid waste by ion chromatography (IC)

    International Nuclear Information System (INIS)

    Chen Lianzhong; Ma Guilan

    1992-01-01

    The determination of Na + and k + ions in the high-level liquid waste is investigated using ion chromatography. In order to protect the low capacity ion exchange resin in single column IC and remove the transition metal as well as other heavy metal ions that are contained in liquid waste, the pretreatment column with EDTA chelating resin is used. Those impurity metal ions are strongly absorbed by EDTA chelating resin and 100% of Na + and K + ions in the solution are eluted. The ability of the decontamination of EDTA chelating resin is satisfactory. The sample of the high-level liquid waste is diluted appropriately, then an aliquot of the sample is passed through the pretreatment column with EDTA chelating resin, the eluate is analysed by single column ion chromatography. The precision of this method is better than 5% for the determination of Na + and K + ions (at μg· ml -1 level)

  11. Innovative processes for the treatment of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Pacary, V.; Barre, Y.; Plasari, E.

    2008-01-01

    Full text of publication follows: Because of the high salinity (0.5 to 2 M) of liquid wastes and the variability of their composition, the method which is the most appropriate and commonly used to remove the contaminants consists in the in situ formation of adsorbent particles in the waste stream. This technique is often called coprecipitation. To increase the efficiency of this treatment, a study is performed to point out the impact of the choice of the process and the influence of operating parameters (mean residence time, stirring speed, etc.) on the formation of crystals and ultimately on their ability to capture radionuclide. Barium sulphate was chosen as a reference because it is a well known precipitate and a material used in the decontamination facilities to remove radiostrontium. Two issues are encountered with the classic treatments which are consequences of the variability of effluents composition. On the one hand when high activity effluents have to be treated, the efficiency of the classic processes can not be sufficient and the liquid must be once again decontaminated. Thus the volume of disposal waste produced by the treatment is doubled. On the other hand when low activity effluents have to be treated, the classic processes produce a low activity waste. Consequently the volume of storage occupied by this waste is disproportionate with regard to its low activity. To return the more flexible process, various configurations were tested. They can be classified in two categories: improvements of the classic treatments and new types of reactors. Because of the good results which are obtained, these processes are patent pending. To support the experimental investigations, a modelling study at the reactor scale is initiated to distinguish the influence of each process parameter. These models assume that the surface of adsorbent particles is continuously renewed by crystal growth. The aim of this work is to determine the decisive parameters which allow the

  12. Removal of fluoride ions from aqueous solution by waste mud

    International Nuclear Information System (INIS)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal; Soylak, Mustafa

    2009-01-01

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy (ΔG o ), enthalpy (ΔH o ), and entropy (ΔS o ) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  13. Removal of fluoride ions from aqueous solution by waste mud

    Energy Technology Data Exchange (ETDEWEB)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal [Karadeniz Technical University, Faculty of Arts and Sciences, Department of Chemistry, 61080 Trabzon (Turkey); Soylak, Mustafa, E-mail: soylak@erciyes.edu.tr [Erciyes University, Faculty of Arts and Sciences, Department of Chemistry, 38039 Kayseri (Turkey)

    2009-09-15

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy ({Delta}G{sup o}), enthalpy ({Delta}H{sup o}), and entropy ({Delta}S{sup o}) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  14. The waste management at research laboratories - problems and solutions

    International Nuclear Information System (INIS)

    Dellamano, Jose Claudio; Vicente, Roberto

    2011-01-01

    The radioactive management in radioactive installations must be planned and controlled. However, in the case of research laboratories, that management is compromised due to the common use of materials and installations, the lack of trained personnel and the nonexistence of clear and objective orientations by the regulator organism. Such failures cause an increasing of generated radioactive wastes and the imprecision or nonexistence of record of radioactive substances, occasioning a financial wastage, and the cancelling of licences for use of radioactive substances. This paper discusses and proposes solutions for the problems found at radioactive waste management in research laboratories

  15. Mixed waste: An alternative solution. The utility perspective

    International Nuclear Information System (INIS)

    Seizert, R.D.

    1988-01-01

    The issue of mixed waste is one of significant interest to the utility industry. The interest is focused on the current regulatory scheme of dual regulation. A fundamental concern of the commercial nuclear utilities resulting from dual regulation is that there are currently no facilities in the US to dispose of mixed low-level radioactive and hazardous waste. The lack of available sites renders mixed waste an orphan, requiring generators of such material to store the waste on-site. This in turn causes commercial nuclear power plants to be subjected to the full gamut of Environmental Protection Agency (EPA) Resource Conservation and Recovery Act (RCRA) regulation in addition to the existing Nuclear Regulatory Commission (NRC) regulations. Superimposing dual regulatory schemes will have impacts which extend far beyond the mere management of mixed waste. Certainly the burdens, complexities and costs of complying with the overlapping regulatory schemes will not have a commensurate increase in protection from the real risks being addressed. For these reasons, the commercial nuclear utility industry is working toward an alternative solution which will protect the public health and the environment from all hazards of mixed waste and will minimize the impacts on both the regulators and the regulated community

  16. Selective extraction and recovery of rare earth metals from phosphor powders in waste fluorescent lamps using an ionic liquid system.

    Science.gov (United States)

    Yang, Fan; Kubota, Fukiko; Baba, Yuzo; Kamiya, Noriho; Goto, Masahiro

    2013-06-15

    The recycling of rare earth metals from phosphor powders in waste fluorescent lamps by solvent extraction using ionic liquids was studied. Acid leaching of rare earth metals from the waste phosphor powder was examined first. Yttrium (Y) and europium (Eu) dissolved readily in the acid solution; however, the leaching of other rare earth metals required substantial energy input. Ionization of target rare earth metals from the waste phosphor powders into the leach solution was critical for their successful recovery. As a high temperature was required for the complete leaching of all rare earth metals, ionic liquids, for which vapor pressure is negligible, were used as an alternative extracting phase to the conventional organic diluent. An extractant, N, N-dioctyldiglycol amic acid (DODGAA), which was recently developed, showed a high affinity for rare earth metal ions in liquid-liquid extraction although a conventional commercial phosphonic extractant did not. An effective recovery of the rare earth metals, Y, Eu, La and Ce, from the metal impurities, Fe, Al and Zn, was achieved from the acidic leach solution of phosphor powders using an ionic liquid containing DODGAA as novel extractant system. Copyright © 2013 Elsevier B.V. All rights reserved.

  17. Qualitative Observations Concerning Packing Densities for Liquids, Solutions, and Random Assemblies of Spheres

    Science.gov (United States)

    Duer, W. C.; And Others

    1977-01-01

    Discusses comparisons of packing densities derived from known molar volume data of liquids and solutions. Suggests further studies for using assemblies of spheres as models for simple liquids and solutions. (MLH)

  18. Operational experience for liquid radioactive waste in FR Yugoslavia

    International Nuclear Information System (INIS)

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2003-01-01

    The present paper reports the results of the preliminary removal of sludge from the bottom of the spent fuel storage pool in the RA reactor, mechanical filtration of the pool water and sludge conditioning and storage. Yugoslavia is a country without a nuclear power plant (NPP) on its territory. The law which strictly forbids NPP construction is still valid, but, nevertheless we must handle and dispose radioactive waste. In the last forty years, in the ''Vinca'' Institute, as a result of two research reactors being operational, named RA and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, radioactive waste materials are stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed in concrete containers, for further manipulation and disposal. (orig.)

  19. Self-aggregation of liquids from biomass in aqueous solution

    International Nuclear Information System (INIS)

    Lomba, Laura; Giner, Beatriz; Zuriaga, Estefanía; Moya, Juana; Lafuente, Carlos

    2013-01-01

    Highlights: • Aggregation behaviour of liquids from biomass in aqueous solution has been studied. • Standard Gibbs free energies of aggregation have been calculated. • Solubility in water of these compounds has been determined. • Critical aggregation concentration decreases as the solubility in water does. -- Abstract: Aggregation of several chemicals from biomass: furfural derived compounds (furfural, 5-methylfurfural, furfuryl alcohol and tetrahydrofurfuryl alcohol), lactate derived compounds (methyl lactate, ethyl lactate and butyl lactate), acrylate derived compound (methyl acrylate) and levulinate compounds (methyl levulinate, ethyl levulinate and butyl levulinate) in aqueous solution has been characterised at T = 298.15 K through density, ρ, speed of sound, u, and isentropic compressibilities, κ S , measurements. In addition the standard Gibbs free energies of aggregation have been also calculated. Furthermore, in order to deepen insight the behaviour of these chemicals in aqueous solution, the solubility of these compounds has been measured at T = 298.15 K

  20. PENGOLAHAN LIMBAH CAIR INDUSTRI SUSU (Liquid Waste Management in Milk Factory

    Directory of Open Access Journals (Sweden)

    Wagini Wagini

    2002-03-01

    Full Text Available ABSTRAK Telah dilakukan suatu penelitian untuk mengetahui kondisi limbah cair industri susu. hasil penelitian menunjukkan bahwa limbah cair industri susu mengandung zat-zat pencemar dalam tingkat yang membahayakan lingkungan, sehingga limbah cair tersebut perlu didaur ulang. Untuk itu diperlukan suatu instalasi peralatan yang mampu mengolah limbah tersebut. Pada penelitian ini proses pengolahan dilakukan dengan mengkombinasikan proses-proses pengolahan secara Fisika, Kimia dan Biologi. Dengan tahapan proses pengolahan yang dipilih meliputi; Proses equalisasi, proses anaerob, proses aerasi, lumpur aktif, proses sedimentasi, proses koagulasi-flokulasi, proses sedimentasi, proses flotasi, proses pengendapan partikel ringan, proses penyaringan dengan pasir dan arang aktif.    Kualitas air hasil pengolahan dianalisa secara Fisika, Kimia dan Biologi melalui parameter-parameter: suhu, kekeruhan, zat padat tersuspensi, zat padat terlarut, daya hantar listrik, PH, BOD, COD dan jumlah bakteri. Penelitian ini menunjukkan air hasil pengolahan aman untuk dibuang ke lingkungan.   ABSTRACT A research to identify the condition of milk industry liquid waste was conducted. The result showed that the waste contained pollutants at the level the endangered the environment. Therefore, the waste had to be recycled in which a liquid waste treatment installation is needed. In this research, the process of milk industry liquid waste was done by combining processing techniques of physics, chemistry and biology. The processing steps include the processes of equalization, anaerobe, aeration, sedimentation, coagulation-flocculation, sedimentation, flotation, sedimentation, filtering with sand and activated carbon. The water resulted from the processes was analyzed in terms of physical, chemical and biological characteristics e.g. temperature, turbidity, suspended solid, solutes solid, conductivity, pH, BOD, COD and amount of bacteria. This research, shows that the water

  1. Molten metal technologies advance waste processing systems for liquid radioactive waste treatment for PWRs and BWRs

    International Nuclear Information System (INIS)

    Strand, Gary; Vance, Jene N.

    1997-01-01

    Molten Metal Technologies (MMT) has recently acquired a proprietary filtration process for specific use in radioactive liquid waste processing systems. The filtration system has been incorporated in to a PWR liquid radwaste system which is currently being designed for the ComEd Byron Nuclear Station. It has also been adopted as the prefiltration step up from of the two RO systems which were part of the VECTRA acquisition and which are currently installed in the ComEd Dresden and Lacily Nuclear Stations. The filtration process has been successfully pilot-tested at both Byron and Dresden and is currently being tested at LaSalle. The important features of the filtration process are the high removal efficiencies for particulates, including colloidal particles, and the low solid waste volume generation per gallon filtered which translates into very small annual solid waste volumes. This filtration process system has been coupled with the use of selective ion exchange media in the PWR processing system to reduce the solid waste volumes generated compared to the current processing methods and to reduce the curie quantities discharged to the environs. In the BWR processing system, this filtration method allows the coupling of an RO system to provide for recycling greater than 95% of the liquid radwaste back to the plant for reuse while significantly reducing the solid waste volumes and operating costs. This paper discusses the process system configurations for the MMT Advanced Waste Processing Systems for both PWRs and BWRs. In addition, the pilot test data and full-scale performance projections for the filtration system are discussed which demonstrate the important features of the filtration process

  2. Treatment of radioactive liquid wastes on semi-permeable membranes

    International Nuclear Information System (INIS)

    Antonescu, M.; Deleanu, N.; Nechifor, G.

    1997-01-01

    At present, among the currently world-wide applied separation processes, those using membranes are thought to be most advanced due to their advantages: high efficiency, cost-effectiveness in application, universality of the utilized equipment, operation in non-destructive and non-polluting conditions. The most significant results of the treatment experiments are: - a reduction of more than 70% in the chemical oxygen consumption for the solution simulating the POD waste; - the solution simulating the secondary waste from decontamination by POD procedure, appear to be the best (with retentions of 88.5%, 76.5% and 65.7% for strontium, cobalt and manganese, respectively). Important reduction of costs and efficient technological schemes can be obtained by combining the semi-permeable membrane separation techniques with other efficient currently used procedures of separation, concentration and purification, adequate for given situations

  3. Extraction of Uranium from Aqueous Solutions Using Ionic Liquid and Supercritical Carbon Dioxide in Conjunction

    International Nuclear Information System (INIS)

    Wang, Joanna S.; Sheaff, Chrystal N.; Yoon, Byunghoon; Addleman, Raymond S.; Wai, Chien M.

    2009-01-01

    Uranyl ions (UO2)2+ in aqueous nitric acid solutions can be extracted into supercritical CO2 (sc-CO2) via an imidazolium-based ionic liquid using tri-n-butylphosphate (TBP) as a complexing agent. The transfer of uranium from the ionic liquid to the supercritical fluid phase was monitored by UV/Vis spectroscopy using a high-pressure fiberoptic cell. The form of the uranyl complex extracted into the supercritical CO2 phase was found to be UO2(NO3)2(TBP)2. The extraction results were confirmed by UV/Vis spectroscopy and by neutron activation analysis. This technique could potentially be used to extract other actinides for applications in the field of nuclear waste management.

  4. Solution processing of polymer semiconductor: Insulator blends-Tailored optical properties through liquid-liquid phase separation control

    KAUST Repository

    Hellmann, Christoph

    2014-12-17

    © 2014 Wiley Periodicals, Inc. It has been demonstrated that the 0-0 absorption transition of poly(3-hexylthiophene) (P3HT) in blends with poly(ethylene oxide) (PEO) could be rationally tuned through the control of the liquid-liquid phase separation process during solution deposition. Pronounced J-like aggregation behavior, characteristic for systems of a low exciton band width, was found for blends where the most pronounced liquid-liquid phase separation occurred in solution, leading to domains of P3HT and PEO of high phase purity. Since liquid-liquid phase separation could be readily manipulated either by the solution temperature, solute concentration, or deposition temperature, to name a few parameters, our findings promise the design from the out-set of semiconductor:insulator architectures of pre-defined properties by manipulation of the interaction parameter between the solutes as well as the respective solute:solvent system using classical polymer science principles.

  5. Solution processing of polymer semiconductor: Insulator blends-Tailored optical properties through liquid-liquid phase separation control

    KAUST Repository

    Hellmann, Christoph; Treat, Neil D.; Scaccabarozzi, Alberto D.; Razzell Hollis, Joseph; Fleischli, Franziska D.; Bannock, James H.; de Mello, John; Michels, Jasper J.; Kim, Ji-Seon; Stingelin, Natalie

    2014-01-01

    © 2014 Wiley Periodicals, Inc. It has been demonstrated that the 0-0 absorption transition of poly(3-hexylthiophene) (P3HT) in blends with poly(ethylene oxide) (PEO) could be rationally tuned through the control of the liquid-liquid phase separation process during solution deposition. Pronounced J-like aggregation behavior, characteristic for systems of a low exciton band width, was found for blends where the most pronounced liquid-liquid phase separation occurred in solution, leading to domains of P3HT and PEO of high phase purity. Since liquid-liquid phase separation could be readily manipulated either by the solution temperature, solute concentration, or deposition temperature, to name a few parameters, our findings promise the design from the out-set of semiconductor:insulator architectures of pre-defined properties by manipulation of the interaction parameter between the solutes as well as the respective solute:solvent system using classical polymer science principles.

  6. Laboratory development of methods for centralized treatment of liquid low-level waste at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Arnold, W.D.; Bostick, D.T.; Burgess, M.W.; Taylor, P.A.; Perona, J.J.; Kent, T.E.

    1994-10-01

    Improved centralized treatment methods are needed in the management of liquid low-level waste (LLLW) at Oak Ridge National Laboratory (ORNL). LLLW, which usually contains radioactive contaminants at concentrations up to millicurie-per-liter levels, has accumulated in underground storage tanks for over 10 years and has reached a volume of over 350,000 gal. These wastes have been collected since 1984 and are a complex mixture of wastes from past nuclear energy research activities. The waste is a highly alkaline 4-5 M NaNO 3 solution with smaller amounts of other salts. This type of waste will continue to be generated as a consequence of future ORNL research programs. Future LLLW (referred to as newly generated LLLW or NGLLLW) is expected to a highly alkaline solution of sodium carbonate and sodium hydroxide with a smaller concentration of sodium nitrate. New treatment facilities are needed to improve the manner in which these wastes are managed. These facilities must be capable of separating and reducing the volume of radioactive contaminants to small stable waste forms. Treated liquids must meet criteria for either discharge to the environment or solidification for onsite disposal. Laboratory testing was performed using simulated waste solutions prepared using the available characterization information as a basis. Testing was conducted to evaluate various methods for selective removal of the major contaminants. The major contaminants requiring removal from Melton Valley Storage Tank liquids are 90 Sr and 137 Cs. Principal contaminants in NGLLLW are 9O Sr, 137 Cs, and 106 Ru. Strontium removal testing began with literature studies and scoping tests with several ion-exchange materials and sorbents

  7. Method of electrolytic processing for radioactive liquid waste

    International Nuclear Information System (INIS)

    Otsuka, Katsuyuki; Takahashi, Yoshiharu; Tamai, Hideaki.

    1989-01-01

    Radioactive liquid wastes containing sodium compounds are electrolized using mercury as a cathode. As a result, they are separated into sodium-containing metal amalgam and residues. Metals containing sodium are separated from amalgam, purified and re-utilized, while mercury is recycled to the electrolysis vessel. The foregoing method can provide advantageous effect such as: (1) volume of the wastes to be processed can be reduced, (2) since processing can be carried out at a relatively low temperature, low boiling elements can be handled with no evaporization, (3) useful elements can be recovered and (4) other method than glass solidification can easily be employed remarkable volume-reduction of solidification products can be expected. (K.M.)

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

    International Nuclear Information System (INIS)

    Anderson, J.D.

    1975-01-01

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

  9. Investigating the effect of compression on solute transport through degrading municipal solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Woodman, N.D., E-mail: n.d.woodman@soton.ac.uk; Rees-White, T.C.; Stringfellow, A.M.; Beaven, R.P.; Hudson, A.P.

    2014-11-15

    Highlights: • The influence of compression on MSW flushing was evaluated using 13 tracer tests. • Compression has little effect on solute diffusion times in MSW. • Lithium tracer was conservative in non-degrading waste but not in degrading waste. • Bromide tracer was conservative, but deuterium was not. - Abstract: The effect of applied compression on the nature of liquid flow and hence the movement of contaminants within municipal solid waste was examined by means of thirteen tracer tests conducted on five separate waste samples. The conservative nature of bromide, lithium and deuterium tracers was evaluated and linked to the presence of degradation in the sample. Lithium and deuterium tracers were non-conservative in the presence of degradation, whereas the bromide remained effectively conservative under all conditions. Solute diffusion times into and out of less mobile blocks of waste were compared for each test under the assumption of dominantly dual-porosity flow. Despite the fact that hydraulic conductivity changed strongly with applied stress, the block diffusion times were found to be much less sensitive to compression. A simple conceptual model, whereby flow is dominated by sub-parallel low permeability obstructions which define predominantly horizontally aligned less mobile zones, is able to explain this result. Compression tends to narrow the gap between the obstructions, but not significantly alter the horizontal length scale. Irrespective of knowledge of the true flow pattern, these results show that simple models of solute flushing from landfill which do not include depth dependent changes in solute transport parameters are justified.

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

    International Nuclear Information System (INIS)

    Amin, E.

    1999-01-01

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

  11. Activated carbons employed to remove ionic liquids from aqueous solutions

    International Nuclear Information System (INIS)

    Hassan, S.; Farooq, A.; Ahmad, M.A.; Irfan, N.; Tufail, M.

    2011-01-01

    Imidazolium and pyridinium based ionic liquids (ILs) have been separated from aqueous solutions by adsorption using a raw Chinese activated carbon (CAC), a bleached Chinese activated carbon (BAC) and an acid treated Chinese activated carbon (AAC) as adsorbent. Adsorption isotherms data of ionic liquids on activated carbons has been obtained. The influence of both cations and anions was analyzed by studying three different ILs. The role of surface chemistry of the adsorbent was also examined using activated carbons modified by oxidative treatments. The BET surface area of activated carbons was measured by nitrogen adsorption. The results of this work indicate that activated carbon is an attractive adsorbent to remove ionic liquids from water streams. It has also been demonstrated that the adsorption of hydrophilic ionic liquids can be improved by modifying the amount and nature of oxygen groups on the activated carbon surface specially by increasing basic groups. The adsorption data for isotherms was studied at acidic, neutral and basic pH values. (author)

  12. Potential agents for removal of actinides from waste solutions

    International Nuclear Information System (INIS)

    Romanovski, V.V.; Whisenhunt, D.W.; Veeck, A.C.; Andersen, W.A.; Hoffman, D.C.; Jide, X.; White, D.; Raymond, K.N.

    1996-01-01

    The uptake of Th(IV) from nitric acid and hydrochloric acid solutions by chelating ion exchange resins containing catechol, 1,2- hydroxypyridinone (1,2-HOPO) and 3,4-hydroxypyridinone (3,4-HOPO) functional groups, has been investigated. These polystyrene based materials show excellent kinetics for uptake of Th(IV) and have a high loading capacity. Liquid/liquid extractants have also been synthesized by addition of lipophilic side chains to the chelating groups (1,2-HOPO; 3,4-HOPO; 3,2-HOPO; catecholamide; terephthalamide). The initial evaluation of the extraction properties has been carried out

  13. Women, e-waste, and technological solutions to climate change.

    Science.gov (United States)

    McAllister, Lucy; Magee, Amanda; Hale, Benjamin

    2014-06-14

    In this paper, we argue that a crossover class of climate change solutions (which we term "technological solutions") may disproportionately and adversely impact some populations over others. We begin by situating our discussion in the wider climate discourse, particularly with regard to the Millennium Development Goals (MDGs) and the Basel Convention. We then suggest that many of the most attractive technological solutions to climate change, such as solar energy and electric car batteries, will likely add to the rapidly growing stream of electronic waste ("e-waste"). This e-waste may have negative downstream effects on otherwise disenfranchised populations. We argue that e-waste burdens women unfairly and disproportionately, affecting their mortality/morbidity and fertility, as well as the development of their children. Building on this, we claim that these injustices are more accurately captured as problems of recognition rather than distribution, since women are often institutionally under-acknowledged both in the workplace and in the home. Without institutional support and representation, women and children are deprived of adequate safety equipment, health precautions, and health insurance. Finally, we return to the question of climate justice in the context of the human right to health and argue for greater inclusion and recognition of women waste workers and other disenfranchised groups in forging future climate agreements. Copyright © 2014 McAllister, Magee. This is an open access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/by-nc/3.0/), which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original author and source are credited.

  14. Advanced evaporation/concentration treatment technology for radioactive liquid waste

    International Nuclear Information System (INIS)

    Zhang Zhijian; Lu Zhiming; Yu Ruixia

    1997-01-01

    A new and effective two stage moisture separator which removes remaining water droplet and free ion in secondary steam can be added between the evaporator and the condenser of existing liquid waste treatment system. Its addition increases decontamination factor to more than ten times. Ion content in condensed water is decreased considerably. Condensed water meets emission standard without passing through ion exchanger. Detail fundamentals are analysed and results are given: (1) system diagram, (2) structure sketch of the two stage moisture separator, (3) laboratory test results

  15. Optimization of a packed bed reactor for liquid waste treatment

    International Nuclear Information System (INIS)

    Schmidt, C.A.; Brower, M.J.; Coogan, J.J.; Tennant, R.A.

    1993-01-01

    The authors describe an optimization study of a packed bed reactor (PBR), developed for the treatment of hazardous liquid wastes. The focus is on the destruction of trichloroethylene (TCE). The PBR technology offers many distinct advantages over other processes: simple design, high destruction rates (99.99%), low costs, ambient pressure operation, easy maintenance and scaleability. The cost effectiveness, optimal operating parameters and scaleability were determined. As a second stage of treatment, a silent discharge plasma (SDP) reactor was installed to further treat offgases from the PBR. A primary advantage of this system is closed loop operation, where exhaust gases are continuously recycled and not released into the atmosphere

  16. Selion offers a unique system for treating liquid nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Tusa, E.; Kurki, H. [ed.

    1998-07-01

    Studies on the treatment of liquid nuclear waste have been conducted actively in the IVO Group since the early 1980s. And the work has borne fruit: the CsTreat and SrTreat ion exchange products, developed by the IVO Group, were launched three years ago. The ion exchangers have already been in full use at a number of sites throughout the world. In addition, they are currently being tested at many nuclear research institutes and power plants in the USA, Japan and Europe

  17. 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.

  18. Disposal of liquid radioactive wastes through wells or shafts

    International Nuclear Information System (INIS)

    Perkins, B.L.

    1982-01-01

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

  19. 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

  20. Uranium,Radium and Iron Absorption from Liquid Waste Uranium Ore Processing by Zeolite

    International Nuclear Information System (INIS)

    Wismawati, T; Sorot sudiro, A; Herjati, T

    1998-01-01

    The aim of this work is to determine zeolites sorption capacity and the distribution coefficient of uranium, radium, and iron in zeolite-liquid waste system. Mineralogical composition of zeolite used in the experiment has been determine by examining the thin sections of zeolite grains under a microscope. Zeolite has ben activated by the dilute sulfuric acid or sodium hydroxide solution. The results show that the use of 0.25 N sodium hydroxide solution could be optimizing the zeolite for uranium and iron ions sorption and that of 0.1 N sulfuric acid solution is for radium sorption. The re-activation process has been carried out in three hours. Under such a condition, the sorption efficiency of zeolite to those ions have been known to be 45.85% for uranium, 96.63 % for iron and 87.80 % for radium. The distribution coefficients of uranium, radium and iron ion in zeolite-liquid waste system have been calculated 0.85, 7.02, and 28.65 ml/g respectively

  1. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

    2006-01-01

    With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge: protection of people and environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, COGEMA LOGISTICS has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear producers through their operational choices. We will focus on the COGEMA LOGISTICS technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfill both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (COGEMA LOGISTICS, TRANSNUCLEAR, MAINCO, and LEMARECHAL CELESTIN) in order to manage transportation of liquid and solid waste towards interim or final storage sites. The paper has the following contents: About radioactive waste; - Radioactive waste classification; - High level activity waste and long-lived intermediate level waste; - Long-lived low level waste; - Short-lived low- and intermediate level waste; - Very low level waste; - The radioactive waste in nuclear fuel cycle; - Packaging design and

  2. An eco friendly solution to the food waste disposal

    Science.gov (United States)

    Babu, G. Reddy; Kumar, G. Madhav

    2017-07-01

    In recent years, waste disposal at workmen camp is one of the major problems being faced by many nations across the world. In the workmen colony at Chittapur, a series of kitchens were built for cooking purpose and a number of small canteens are also functioning. Considerable quantity of food waste is collected daily from these eateries and disposed at a faraway place. Food waste is highly degradable in nature, if not disposed properly it causes problems related to environmental pollution. Hence, it is very important to identify an environment friendly process rather than opt for land filling or any disposal method. We worked together to find a suitable eco-friendly solution for the food waste disposal at Chittapur site and suggested that biogas production through anaerobic digestion is a solution for the disposal and utilization of food waste for better purpose. This resulted in setting up a 500 kg per day food waste treatment biogas plant at Chittapur. This establishment is the first time in the construction industry at workmen camp in India. Anaerobic Digestion has been recognized as one of the best options that is available for treating food waste, as it generates two valuable end products, biogas and compost. Biogas is a mixture of CH4 and CO2 about (55:45). Biogas generated can be used for thermal applications such as cooking or for generating electricity. The digested slurry is a well stabilized organic manure and can be used as soil fertilizer. Plant design is to handle 500 kg of food waste /day. 27 kg LPG is obtained from 500kg of kitchen waste. The Value of 27 kg of LPG is Rs.2700/day. Daily 1000 litres of digested effluent was obtained. It is good organic manure with plant micro nutrients and macro nutrients. This can be used for growing plants and in agriculture. The value of manure per day is Rs.250/-. The annual revenue is Rs.10.62 lakhs and the annual expenditure is 1.8 lakhs. The net benefit is 8.82 lakhs. Payback period is 2.1 years. This process

  3. Modeling of the (liquid + liquid) equilibrium of polydisperse hyperbranched polymer solutions by lattice-cluster theory

    International Nuclear Information System (INIS)

    Enders, Sabine; Browarzik, Dieter

    2014-01-01

    Graphical abstract: - Highlights: • Calculation of the (liquid + liquid) equilibrium of hyperbranched polymer solutions. • Description of branching effects by the lattice-cluster theory. • Consideration of self- and cross association by chemical association models. • Treatment of the molar-mass polydispersity by the use of continuous thermodynamics. • Improvement of the theoretical results by the incorporation of polydispersity. - Abstract: The (liquid + liquid) equilibrium of solutions of hyperbranched polymers of the Boltorn type is modeled in the framework of lattice-cluster theory. The association effects are described by the chemical association models CALM (for self association) and ECALM (for cross association). For the first time the molar mass polydispersity of the hyperbranched polymers is taken into account. For this purpose continuous thermodynamics is applied. Because the segment-molar excess Gibbs free energy depends on the number average of the segment number of the polymer the treatment is more general than in previous papers on continuous thermodynamics. The polydispersity is described by a generalized Schulz–Flory distribution. The calculation of the cloud-point curve reduces to two equations that have to be numerically solved. Conditions for the calculation of the spinodal curve and of the critical point are derived. The calculated results are compared to experimental data taken from the literature. For Boltorn solutions in non-polar solvents the polydispersity influence is small. In all other of the considered cases polydispersity influences the (liquid + liquid) equilibrium considerably. However, association and polydispersity influence phase equilibrium in a complex manner. Taking polydispersity into account the accuracy of the calculations is improved, especially, in the diluted region

  4. Distillation as a pretreatment process of waste scintillation solutions

    International Nuclear Information System (INIS)

    Dellamano, J.C.

    1988-05-01

    A process to pretreat scintillation solutions composed basically of PPO, POPOP, TOLUENE and ANTAROX, utilized by radioimmunoassay laboratories, is described. The technique employed is distillation which permits a waste reduction to about 40% of the initial volume with the recovery of the solvent (toluene). The recovered toluene can be resued for the same purpose, since it is free of radioactive material as assured by quality control procedures. (author) [pt

  5. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

  6. 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.

  7. Evaluation of extractant-coated magnetic microparticles for the recovery of hazardous metals from waste solution

    International Nuclear Information System (INIS)

    Kaminski, M. D.

    1998-01-01

    A magnetically assisted chemical separation (MACS) process was developed earlier at Argonne National Laboratory (ANL). This compact process was designed for the separation of transuranics (TRU) and radionuclides from the liquid waste streams that exist at many DOE sites, with an overall reduction in waste volume requiring disposal. The MACS process combines the selectivity afforded by solvent extractant/ion exchange materials with magnetic separation to provide an efficient chemical separation. Recently, the MACS process has been evaluated with acidic organophosphorus extractants for hazardous metal recovery from waste solutions. Moreover, process scale-up design issues have been addressed with respect to particle filtration and recovery. Two acidic organophosphorus compounds have been investigated for hazardous metal recovery, bis(2,4,4-trimethylpentyl) phosphinic acid (Cyanexreg-sign 272) and bis(2,4,4-trimethylpentyl) dithiophosphinic acid (Cyanexreg-sign 301). Coated onto magnetic microparticles, these extractants demonstrated superior recovery of hazardous metals from solution, relative to what was expected on the basis of results from solvent extraction experiments. The results illustrate the diverse applications of MACS technology for dilute waste streams. Preliminary process scale-up experiments with a high-gradient magnetic separator at Oak Ridge National Laboratory have revealed that very low microparticle loss rates are possible

  8. Extraction of zirconium from simulated acidic nitrate waste using liquid membrane in hollow fiber contactor

    International Nuclear Information System (INIS)

    Pandey, G.; Chinchale, R.; Renjith, A.U.; Dixit, S.; Mukhopadhyay, S.; Shenoy, K.T.; Ghosh, S.K.

    2015-01-01

    The acidic waste raffinate stream of zirconium (Zr) purification plant contains about 2 gpl of Zr in about 2M free nitric acid. TBP, which is the most commonly used solvent in the nuclear industry, is not suitable for the extraction of Zr from this lean solution as its distribution coefficient is less than one. In house synthesized Mixed Alkyl Phosphine Oxide (MAPO) is a potential extractant for Zr from this lean stream. Intensification of this process for recovery of Zr has been attempted through use of efficient contactor, namely, hollow fiber module and efficient process, namely, simultaneous extraction and stripping across liquid membrane containing MAPO. Based on batch equilibrium studies selection of suitable concentration of extractant, composition of diluent, selection and concentration of strippant for the proposed liquid membrane system was made. The selected organic and strippant concentration was used to study suitability of application of Dispersion Liquid Membrane (DLM) in hollow fiber contactor for recovery Zr from solution simulated to Zr plant raffinate. Challenges related to stable operation of the liquid membrane system like stability of the organic phase in the micropores of lumen and stability of the dispersion during the pertraction were addressed through pressure balance across the lumen and choice of adequate dispersion condition respectively. (author)

  9. Engineering study radioactive liquid waste treatment plant refurbishment

    International Nuclear Information System (INIS)

    Suazo, I.L.

    1994-01-01

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

  10. Chemical treatment of radioactive liquid wastes from medical applications

    International Nuclear Information System (INIS)

    Castillo A, J.

    1995-01-01

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

  11. Poly (ether imide sulfone) membranes from solutions in ionic liquids

    KAUST Repository

    Kim, Dooli

    2017-11-20

    A membrane manufacture method based on non-volatile solvents and a high performance polymer, poly (ether imide sulfone) (EXTEM™), is proposed, as greener alternative to currently industrial process. We dissolved EXTEM™ in pure ionic liquids: 1-ethyl-3-methylimidalzolium thiocyanate ([EMIM]SCN), 1-butyl-3-methylimidalzolium thiocyanate ([BMIM]SCN), and 1-ethyl-3-methylimidalzolium acetate ([EMIM]OAc). The following polymer solution parameters were evaluated to optimize the manufacture: Gibbs free energy of mixing (G), intrinsic viscosity ([]) and hydrodynamic diameter. Membranes with sponge-like structure and narrow pore size distribution were obtained from solutions in [EMIM]SCN. They were tested for separation of proteins and deoxyribonucleic acids (DNA). Due to the polymer stability, we foresee that applications in more demanding chemical separations would be possible. [EMIM]SCN was 96 % purified and recovered after the membrane fabrication, contributing to the sustainability of the whole manufacturing process.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  13. Modeling the liquid-liquid interface and the transfer of a solute by molecular dynamics simulation

    International Nuclear Information System (INIS)

    Hayoun, Marc

    1990-11-01

    Molecular Dynamics method and Lennard-Jones potential functions have been employed to model Liquid-Liquid Interfaces. The variation of the miscibilities between the two liquids is obtained by changing the interaction between the two atomic species. The resulting interfaces have a thickness of about three atomic diameters and are stable on the time scale of the simulation. They have been characterized by the density and pressure profiles. The interfacial tension has also been computed and is of the order of magnitude of experimental values. The diffusion process is anisotropic in the interfacial region: the transverse diffusion coefficient (parallelly to the interface) is higher than the normal one. A qualitative explanation of this behaviour is suggested by considering the pressure tensor. The second part of this work, performed by Molecular Dynamics in the canonical ensemble, is devoted to the kinetic study of the transfer of a solute through the interface. A model of a symmetric interface with an atomic solute has been used. The interaction potential between the solute and the solvents has been built in order to obtain an activation barrier to the transfer. We have computed the mean force exerted by the solvent on the solute as a function of its distance to the interface. The resulting mean force potential corresponds to a free energy difference. The height of the energy barrier involved is about 4 kT. The potential energy and entropy profiles have also been calculated and discussed. The diffusion coefficient of the solute has been computed by equilibrium and non-equilibrium methods. We deduced the friction coefficient of the solvent, which is essential to determine the Kramers transmission coefficient. This coefficient is compared to the one obtained by simulation. Finally, the solute transfer rate constant has been calculated. (author) [fr

  14. Separation and recovery of sodium nitrate from low-level radioactive liquid waste by electrodialysis

    International Nuclear Information System (INIS)

    Meguro, Yoshihiro; Kato, Atsushi; Watanabe, Yoko; Takahashi, Kuniaki

    2011-01-01

    An advanced method, in which electrodialysis separation of sodium nitrate and decomposition of nitrate ion are combined, has been developed to remove nitrate ion from low-level radioactive liquid wastes including nitrate salts of high concentration. In the electrodialysis separation, the sodium nitrate was recovered as nitric acid and sodium hydroxide. When they are reused, it is necessary to reduce the quantity of impurities getting mixed with them from the waste fluid as much as possible. In this study, therefore, a cation exchange membrane with permselectivity for sodium ion and an anion exchange membrane with permselectivity for monovalent anion were employed. Using these membranes sodium and nitrate ions were effectively removed form a sodium nitrate solution of high concentration. And also it was confirmed that sodium ion was successfully separated from cesium and strontium ions and that nitrate ion was separated from sulfate and phosphate ions. (author)

  15. Possible use of ionic polymers for treatment of radioactive liquid waste

    International Nuclear Information System (INIS)

    Siyam, T.; Nofal, M.; Eldessouky, M.I.; Aly, H.F.

    1992-01-01

    Water-soluble nonionic polymers such as polyacrylamide is recently introduced for treatment of radioactive liquid waste. Eater-soluble ionic polymers such as: poly (sodium acrylate) [anionic polymer], poly (acrylamide-CO-sodium acrylate) [anionic copolymer] and poly (acrylamide-sodium acrylate-diallyldiethylammonium chloride) [amphoteric terpolymer] were prepared by gamma radiation-initiated polymerization of the corresponding monomer solutions. The prepared polymers were assessed for use in treatment of radionuclides that might be present in radioactive waste effluents. It was found that the polymer efficiency for cobalt-60 was affected by the composition of the copolymer and the degree of ionization of the polymer. The efficiency of the polymer increases with increasing the concentration of the polymer. The mechanism of sludge formation for each type of polymer was discussed. It was found that the anionic copolymer is more selective for cobalt than the prepared polymers. Amphoteric terpolymer has different selectivity for cations and anions. 3 figs, 1 tab

  16. A liquid-liquid transition in supercooled aqueous solution related to the HDA-LDA transition

    Science.gov (United States)

    Woutersen, Sander; Ensing, Bernd; Hilbers, Michiel; Zhao, Zuofeng; Angell, C. Austen

    2018-03-01

    Simulations and theory suggest that the thermodynamic anomalies of water may be related to a phase transition between two supercooled liquid states, but so far this phase transition has not been observed experimentally because of preemptive ice crystallization. We used calorimetry, infrared spectroscopy, and molecular dynamics simulations to investigate a water-rich hydrazinium trifluoroacetate solution in which the local hydrogen bond structure surrounding a water molecule resembles that in neat water at elevated pressure, but which does not crystallize upon cooling. Instead, this solution underwent a sharp, reversible phase transition between two homogeneous liquid states. The hydrogen-bond structures of these two states are similar to those established for high- and low-density amorphous (HDA and LDA) water. Such structural similarity supports theories that predict a similar sharp transition in pure water under pressure if ice crystallization could be suppressed.

  17. Determination of difficult to measure actinides in radioactive liquid waste

    International Nuclear Information System (INIS)

    Drabova, V.; Galanda, D.; Dulanska, S.; Remenec, B.; Kuruc, J.

    2014-01-01

    In decommissioning of a nuclear facilities and radioactive waste treatment the activity of various radionuclides need to be measured for the waste characterization. Radiochemical separation of 241 Am, 237 Np and isotopes of plutonium was tested on model solution of evaporator concentrate sample for isolation of each of them for alpha-spectrometry analysis. This paper describes use of the molecular recognition technology product AnaLig(R)Pu-01 gel from IBC Advanced technologies, Inc. to effectively and selectively pre-concentrate, separate and recover difficult-to-measure actinides from model solution of evaporator concentrate samples which belong to the most difficult matrices to analyse. The method is suitable for analysing highly contaminated samples of radioactive waste in a relatively short time. For counting the alpha activity of 241 Am, 239,240 Pu, 238 Pu and 237 Np ORTEC 576A alpha-spectrometer equipped with ULTRA TM ion implanted silicon detectors (600 mm 2 active area) was used. The spectra were processed by using the Alpha-vision TM 32-bit emulation software from the EG and G ORTEC company. (authors)

  18. Geopolymerisation of fly ashes with waste aluminium anodising etching solutions.

    Science.gov (United States)

    Ogundiran, M B; Nugteren, H W; Witkamp, G J

    2016-10-01

    Combined management of coal combustion fly ash and waste aluminium anodising etching solutions using geopolymerisation presents economic and environmental benefits. The possibility of using waste aluminium anodising etching solution (AES) as activator to produce fly ash geopolymers in place of the commonly used silicate solutions was explored in this study. Geopolymerisation capacities of five European fly ashes with AES and the leaching of elements from their corresponding geopolymers were studied. Conventional commercial potassium silicate activator-based geopolymers were used as a reference. The geopolymers produced were subjected to physical, mechanical and leaching tests. The leaching of elements was tested on 28 days cured and crushed geopolymers using NEN 12457-4, NEN 7375, SPLP and TCLP leaching tests. After 28 days ambient curing, the geopolymers based on the etching solution activator showed compressive strength values between 51 and 84 MPa, whereas the commercial potassium silicate based geopolymers gave compressive strength values between 89 and 115 MPa. Based on the regulatory limits currently associated with the used leaching tests, all except one of the produced geopolymers (with above threshold leaching of As and Se) passed the recommended limits. The AES-geopolymer geopolymers demonstrated excellent compressive strength, although less than geopolymers made from commercial activator. Additionally, they demonstrated low element leaching potentials and therefore can be suitable for use in construction works. Copyright © 2016. Published by Elsevier Ltd.

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

    International Nuclear Information System (INIS)

    Lee, Soonmin; Kim, Kiljung; Park, Jungsu

    2008-01-01

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

  20. Treatment of low radioactive liquid waste by electrodialysis. Principles and experimental model

    International Nuclear Information System (INIS)

    Dogaru, D.

    1998-01-01

    Electrodialysis is a membrane separation process achieved by the use of differential driving force due to an electric potential across the membrane. It can be considered as a process in which salts are transferred under the impetus of an electrical potential from one solution to another, usually from a dilute to a concentrated solution, through a membrane barrier. In water, salts dissolve producing positively charged cations and negatively charged anions. If an electrical field is placed across a solution of salt by inserting a pair of electrodes into the solution, the cations migrate toward the negatively charged cathode, while anions migrate toward the positively charged anode. This contribution presents principles and experimental model for removed radionuclides from low radioactive liquid wastes. A typical electrodialysis cell arrangement consists of a series of anionic- and cationic-exchange membranes arranged in an alternating pattern between an anode and a cathode, to form an individual cell. The laboratory experimental apparatus consisted of an electrodialysis unit, two recirculating pumps, a voltage stabilizer, connecting pipes and recirculating tanks. The unit had 10 cell pairs. The cell geometry was a flat-plate and frame configuration with anode, cathode, charge selective membranes, gaskets and spacers. The anode material was nickel and the cathode material was TiO 2 , with an electrode area of 90 cm 2 . For Radioactive Waste Treatment Plant, the ability to separate equivalent ions is very attractive and opens the possibility of applying electrodialysis to a wide variety of systems with appropriate choice of operating conditions and ion-selective membranes. The technique creates minimal secondary waste. However, before electrodialysis can be implemented, a chemical pre-treatment for radioactive wastes is necessary. (author)

  1. Formation and filtration characteristics of solids generated in a high level liquid waste treatment process. Solids formation behavior from simulated high level liquid waste

    International Nuclear Information System (INIS)

    Kondo, Y.; Kubota, M.

    1997-01-01

    The solids formation behavior in a simulated high level liquid waste (HLLW) was experimentally examined, when the simulated HLLW was treated in the ordinary way of actual HLLW treatment process. Solids formation conditions and mechanism were closely discussed. The solids formation during a concentration step can be explained by considering the formation of zirconium phosphate, phosphomolybdic acid and precipitation of strontium and barium nitrates and their solubilities. For the solids formation during the denitration step, at least four courses were observed; formation of an undissolved material by a chemical reaction with each other of solute elements (zirconium, molybdenum, tellurium) precipitation by reduction (platinum group metals) formation of hydroxide or carbonate compounds (chromium, neodymium, iron, nickel, strontium, barium) and a physical adsorption to stable solid such as zirconium molybdate (nickel, strontium, barium). (author)

  2. Uranium Extraction From Artificial Liquid Waste Using Continuous Extraction Liquid membrane Technique

    International Nuclear Information System (INIS)

    Rusdianasari; Buchari

    2002-01-01

    The continuous extraction of uranium from artificial liquid waste by emulsion liquid membrane was carried out using one stage mixer-settler. This emulsion liquid membrane containing di-2-ethylhexylphosphoric acid (D2EHPA) and tri-n-buthyl phosphate (TBP) as carrier were carried out using one stage mixer-settler. The optimum condition gave the ratio of emulsion velocity to the feed velocity 1:4 and steady state reached after five minutes. The optimum condition was obtained at the 90.91 % of uranium recovered from raffinate, using EDTA as the masking agent with concentration 5x10 - 2 M . The total concentration of carrier was 3% with ratio D2EHPA and TBP 3:1. The emulsion liquid membrane has high relative selectivity after steady state with separation factors were α U , N i= 115,43 and α U , Fe 328,55. The result of experiment showed that emulsion liquid membrane containing D2EHPA and TBP as carrier have good performance for continuous system

  3. Evaporation studies on Oak Ridge National Laboratory liquid low-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, V.L. [PAI Corp., Oak Ridge, TN (United States); Perona, J.J. [Oak Ridge National Lab., TN (United States)

    1993-03-01

    Evaporation studies were performed with Melton Valley storage tank liquid low-level radioactive waste concentrate and with surrogates (nonradioactive) to determine the feasibility of a proposed out-of-tank-evaporation project. Bench-scale tests indicated that volume reductions ranging from 30 to 55% could be attained. Vendor-site tests were conducted (with surrogate waste forms) using a bench-scale single-stage, low-pressure (subatmospheric), low-temperature (120 to 173{degree}F) evaporator similar to units in operation at several nuclear facilities. Vendor tests were successful; a 30% volume reduction was attained with no crystallization of solids and no foaming, as would be expected from a high pH solution. No fouling of the heat exchanger surfaces occurred during these tests. It is projected that 52,000 to 120,000 gal of water could be evaporated from the supernate stored in the Melton and Bethel Valley liquid low-level radioactive waste (LLLW) storage tanks with this type of evaporator.

  4. PMP, a novel solute for liquid and plastic scintillation counting

    International Nuclear Information System (INIS)

    Gusten, Hans

    1983-01-01

    The excellent fluorescence properties of PMP ( 11-phenyl-3-mesityl-2-pyrazoline) such as long wavelength emission of over 400 nm, and high fluorescence quantum yield with a short decay time together with a solubility of more than one Mol/L in toluene make this compound a promising solute for scintillation counting. The Stokes' shift of PMP of over 10,000 cm -1 is twice as large as that of the commonly used PPO. Due to this unusually large Stokes' shift PMP can be used as a primary solute without requiring a secondary solute as wavelength shifter. A comparison of the scintillation properties of PMP and PPO in toluene reveals that the counting efficiency for 14 C is better for PMP while the 3 H efficiency is equally good. Due to the large Stokes' shift, PMP is about 50 percent less sensitive to color quenching than PPO. Compared to the solute combinations PPO/secondary solutes, the scintillation counting efficiency of PMP for 14 C in toluene or xylene is the same, while the absolute 3 H efficiency of PPO/secondary solutes in cocktails with emulsifiers is about 10 percent higher. The PMP scintillation efficiency for 14 C as well as 3 H in chemical quenching by urine is more or less the same as for PPO/dimethyl-POPOP. PMP is more sensitive to quenching by halogenated solvents. In the dioxane-based scintillation, this sensitivity to chemical quenching by CHCl 3 vanishes and the counting efficiencies for 14 C and 3 H are as good as for PPO/dimethyl-POPOP or PPO/bis-MSB. Due to the large Stokes' shift, the self-absorption of the scintillation light by PMP is lower than in conventional scintillators. This offers good possibilities in very large-volume applications of liquid as well as plastic scintillators

  5. Development of Concentration and Calcination Technology For High Level Liquid Waste

    International Nuclear Information System (INIS)

    Pande, D.P.

    2006-01-01

    The concentrated medium and high-level liquid radio chemicals effluents contain nitric acid, water along with the dissolved chemicals including the nitrates of the radio nuclides. High level liquid waste contain mainly nitrates of cesium, strontium, cerium, zirconium, chromium, barium, calcium, cobalt, copper, pickle, iron etc. and other fission products. This concentrated solution requires further evaporation, dehydration, drying and decomposition in temperature range of 150 to 700 deg. C. The addition of the calcined solids in vitrification pot, instead of liquid feed, helps to avoid low temperature zone because the vaporization of the liquid and decomposition of nitrates do not take place inside the melter. In our work Differential and thermo gravimetric studies has been carried out in the various stages of thermal treatment including drying, dehydration and conversion to oxide forms. Experimental studies were done to characterize the chemicals present in high-level radioactive waste. A Rotary Ball Kiln Calciner was used for development of the process because this is amenable for continuous operation and moderately good heat transfer can be achieved inside the kiln. This also has minimum secondary waste and off gases generation. The Rotary Ball Kiln Calciner Demonstration facility system was designed and installed for the demonstration of calcination process. The Rotary Ball Kiln Calciner is a slowly rotating slightly inclined horizontal tube that is externally heated by means of electric resistance heating. The liquid feed is sprayed onto the moving bed of metal balls in a slowly rotating calciner by a peristaltic type-metering pump. The vaporization of the liquid occurs in the pre-calcination zone due to counter current flow of hot gases. The dehydration and denitration of the solids occurs in the calcination zone, which is externally heated by electrical furnace. The calcined powder is cooled in the post calcination portion. It has been demonstrated that the

  6. 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)

  7. Technical experiences for liquid radioactive waste in FR Yugoslavia

    International Nuclear Information System (INIS)

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2002-01-01

    Yugoslavia is a country without any Nuclear Power Plant on its territory. In the last forty years, in Nuclear Sciences Institute 'Vinca', as a result of the two reactors operation, named RA and RB, and as a result of the radionuclides application in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, radioactive waste materials are stored in two interim storage facilities. Radwaste that were immobilized in the inactive matrices are to be placed into the concrete containers, for the further manipulation and disposal. The present paper reports the results on preliminary removal of sludge from the bottom of the spent fuel storage pool in RA reactor, mechanical filtration of the pool water and sludge conditioning and storage. (authors)

  8. Removal of palladium precipitate from a simulated high-level radioactive liquid waste by reduction by ascorbic acid

    International Nuclear Information System (INIS)

    Kim, Eung Ho; Yoo, Jae Hyung; Choi, Cheong Song

    1998-01-01

    A study of the selective removal of Palladium from a simulated solution of high-level radioactive liquid waste (HLLW) was carried out. The simulated solution contained 7 representative elements (Pd 2+ , Cs + , Sr 2+ , Fe 3+ , MoO 2 2+ , Ru 4+ , and Nd 3+ ) typical of HLLW, ascorbic acid was added to the solution at room temperature. Pd 2+ in the simulated solution was easily reduced to Pd metal by the ascorbic acid and then the metal precipitate could be removed from the solution, whereas other elements remained mainly in solution. When the resulting Pd metal was left in solution, it was reoxidized to Pb 2+ ion and redissolved in a nitric acid medium. The oxidation rate of Pd 2+ depended on the presence of a transition metal such as ferric ion, and was also in proportion to the concentration of nitric acid and in inverse proportion to the concentration of ascrobic acid. (orig.)

  9. 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.)

  10. 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.)

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

    International Nuclear Information System (INIS)

    Kong Jinsong; Guo Weiqun; Lu Jingbin

    2012-01-01

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

  12. Decontamination of liquid radioactive wastes using seeded ultrafiltration

    International Nuclear Information System (INIS)

    Kavanagh, P.; Goldsmith, A.

    1997-01-01

    A number of techniques may be used to treat radioactive wastes. This paper presents a discussion of the relative merits of two of these: ion exchange and membrane filtration, and discusses the overall benefit of using seeded filtration to combine the advantages of each, with selected examples of where these techniques have been used. Evaporation is another technique that can be used, however, because of its high capital and operating costs its use is limited and it is not discussed here. Examples of the decontamination of standard solutions by novel materials tested by the Novel Absorber Evaluation Club are presented, and the advantages of the new PAN-based absorbers discussed

  13. Mercury removal from liquid and solid mixed waste

    International Nuclear Information System (INIS)

    Gates, D.D.; Klasson, K.T.; Corder, S.L.; Cameron, P.A.; Perona, J.J.

    1995-01-01

    Based on bench-scale laboratory experiments, the following conclusions were reached: Sulfur-impregnated, activated, carbon pellets (Mersorb) can be used to remove mercury (Hg 2+ ) to below EPA's toxic characteristic level (0.2 mg/L). Mersorb works under acid conditions (pH 2) but its capacity is reduced by approximately 50% compared with neutral conditions. Competing ions present in the target waste stream reduced the Mersorb capacity by 50%. Mersorb appears to be economical compared with leading ion exchange resin. KI/I 2 leaching solution can be used to remove up to 99% of Hg in contaminated soil and glass. KI/I 2 leaching solution worked well with several mercury species, including Hg 0 , HgO, HgS, and HgCl 2 . KI/I 2 leaching solution worked well with a wide variety of initial mercury concentrations. Radionuclide surrogate studies suggested that uranium will not partition into KI/I 2 leaching solutions. Cesium may partition into the KI/I 2 leaching solution because of the high solubility of cesium salts

  14. Measurement of Solute Diffusion Behavior in Fractured Waste Glass Media

    International Nuclear Information System (INIS)

    Saripalli, Kanaka P.; Lindberg, Michael J.; Meyer, Philip D.

    2008-01-01

    Determination of aqueous phase diffusion coefficients of solutes through fractured media is essential for understanding and modeling contaminants transport at many hazardous waste disposal sites. No methods for earlier measurements are available for the characterization of diffusion in fractured glass blocks. We report here the use of time-lag diffusion experimental method to assess the diffusion behavior of three different solutes (Cs, Sr and Pentafluoro Benzoic Acid or PFBA) in fractured, immobilized low activity waste (ILAW) glass forms. A fractured media time-lag diffusion experimental apparatus that allows the measurement of diffusion coefficients has been designed and built for this purpose. Use of time-lag diffusion method, a considerably easier experimental method than the other available methods, was not previously demonstrated for measuring diffusion in any fractured media. Hydraulic conductivity, porosity and diffusion coefficients of a solute were experimentally measured in fractured glass blocks using this method for the first time. Results agree with the range of properties reported for similar rock media earlier, indicating that the time-lag experimental method can effectively characterize the diffusion coefficients of fractured ILAW glass media

  15. Industrial solid and liquid waste treatment processes; Les procedes de traitement des dechets industriels solides et liquides

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1995-11-01

    This catalogue gives information on 68 chemical, mechanical, magnetic, electrical, thermal, etc. techniques for the processing of solid, viscous and liquid, common or special, industrial wastes. The various processes are presented as files, which are easily retrievable through keywords, waste type or industry codes, processing types, distributors. Technologies, performances and applications of each techniques are presented, together with references and company contacts

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  17. Determination of service standard time for liquid waste parameter in certification institution

    Science.gov (United States)

    Sembiring, M. T.; Kusumawaty, D.

    2018-02-01

    Baristand Industry Medan is a technical implementation unit under the Industrial and Research and Development Agency, the Ministry of Industry. One of the services often used in Baristand Industry Medan is liquid waste testing service. The company set the standard of service 9 working days for testing services. At 2015, 89.66% on testing services liquid waste does not meet the specified standard of services company. The purpose of this research is to specify the standard time of each parameter in testing services liquid waste. The method used is the stopwatch time study. There are 45 test parameters in liquid waste laboratory. The measurement of the time done 4 samples per test parameters using the stopwatch. From the measurement results obtained standard time that the standard Minimum Service test of liquid waste is 13 working days if there is testing E. coli.

  18. Decontamination of radioactive liquid wastes by hydrophytic vegetal organisms

    International Nuclear Information System (INIS)

    Cecal, Al; Popa, K.; Potoroaca, V.; Melniciuc-Puica, N.

    2001-01-01

    Bioaccumulation of some radioactive ions from contaminated waste solutions, on hydrophytic vegetal organisms is discussed. In order to follow the distribution of radioactive ions 137 Cs + , 60 Co 2+ and 51 Cr 3+ in various cell components extracted from Spirulina platensis, Porphiridium cruentum, Scenedesmus quadricauda, Lemna minor, Elodea canadensis, Pistia stratiotes and Riccia fluitans, the plants were cultivated in radioactive solutions. The resulting complexes were extracted with acetone or acetic acid and separated chromatographically. The results show an intense activity of the polysaccharide and lipoid fractions in the bioaccumulation process. The bioaccumulation varies in the series: Spirulina > Scenedesmus > Porphiridium > Riccia > Pistia > Lemna ≥ Elodea for 137 Cs + and 60 Co 2+ ; Spirulina > Porphiridium > Scenedesmus > Riccia > Pistia > Lemna > Elodea for 51 Cr 3+ . (author)

  19. Inhibition of nuclear waste solutions containing multiple aggressive anions

    International Nuclear Information System (INIS)

    Congdon, J.W.

    1987-01-01

    The inhibition of localized corrosion of carbon steel in caustic, high-level radioactive waste solutions was studied using cyclic potentiodynamic polarization scans, supplemented by partially immersed coupon tests. The electrochemical tests provided a rapid and accurate means of determining the relationship between the minimum inhibitor requirements and the concentration of the aggressive anions in this system. Nitrate, sulfate, chloride, and fluoride were identified as aggressive anions, however, no synergistic effects were observed between these anions. This observation may have important theoretical implications because it tends to contradict the behavior of aggressive anions as predicted by existing theories for localized corrosion. 10 refs., 5 figs., 2 tabs

  20. Electrochemical ion-exchange for medium active liquid waste treatment

    International Nuclear Information System (INIS)

    Bridger, N.J.; Turner, A.D.

    1987-01-01

    Electrochemical ion-exchange has already been demonstrated to be a robust, effective process for the treatment of active liquid wastes -with high decontamination and volume reduction factors, and only a low energy requirement. The primary aim of this new programme is to scale up this process - initially to 0.1m 3 /h, and ultimately to 1 3 m/h. A new 0.4m 2 electrode module has been designed and constructed, together with 3m 3 feed tanks for the first phase of this work. Further development work is also being carried out on alternative electrode designs and fabrication methods, as well as new exchange media (including inorganic absorbers and organic chelating resins) in order to optimize selectivity performance. (author)

  1. Low-level liquid waste decontamination by organic ion exchange

    International Nuclear Information System (INIS)

    Lee, D.D.; Campbell, D.O.; Dillow, T.A.

    1990-01-01

    Improved processes are being developed to treat contaminated liquid wastes that have been and continue to be generated at Oak Ridge National Laboratory. Promising results have been obtained for cesium removal with a new resorcinol-based organic resin developed at the Savannah River Site. In tests of cesium removal, it was superior to other available resins, such as Duolite CS-100, with the distribution coefficient being limited primarily by competition from potassium and nearly independent of the sodium concentration. The optimum pH was approximately 12.5 in high NaNO 3 concentrations (>2 M). A fairly low flow velocity was required to yield sharp breakthrough of the loaded cesium. The resin was much less effective for strontium removal, which was limited by competition from sodium. If both cesium and strontium must be removed, another resin column or a mixed bed with a chelating resin should be used

  2. Radioactive waste and special waste disposal in salt domes - phoney waste management solutions

    International Nuclear Information System (INIS)

    Grimmel, E.

    1990-01-01

    The paper tries to make aware of the fact that an indefinite safe disposal of anthropogeneous wastes in underground repositories is impossible. Suspicion is raised that the Gorleben-Rambow salt dome has never been studied for its suitability as a repository, but that it was simply taken for granted. Safety analyses are meant only to conceal uncertainty. It is demanded to immediately opt out of the ultimate disposal technique for radioactive and special wastes in salt caverns. (DG) [de

  3. Photometric estimation of plutonium in product solutions and acid waste solutions using flow injection analysis technique

    International Nuclear Information System (INIS)

    Dhas, A.J.A.; Dharmapurikar, G.R.; Kumaraguru, K.; Vijayan, K.; Kapoor, S.C.; Ramanujam, A.

    1995-01-01

    Flow injection analysis technique is employed for the measurement of plutonium concentrations in product nitrate solutions by measuring the absorbance of Pu(III) at 565 nm and of Pu(IV) at 470 nm, using a Metrohm 662 photometer, with a pyrex glass tube of 2 nm (ID) inserted in the light path of the detector serving as a flow cell. The photometer detector never comes in contact with radioactive solution. In the case of acid waste solutions Pu is first purified by extraction chromatography with 2-ethyl hexyl hydrogen 2 ethyl hexyl phosphonate (KSM 17)- chromosorb and the Pu in the eluate in complexed with Arsenazo III followed by the measured of absorbance at 665 nm. Absorbance of reference solutions in the desired concentration ranges are measured to calibrate the system. The results obtained agree with the reference values within ±2.0%. (author). 3 refs., 1 tab

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  5. Community Solutions to Solid Waste Pollution. Operation Waste Watch: The New Three Rs for Elementary School. Grade 6. [Second Edition.

    Science.gov (United States)

    Virginia State Dept. of Waste Management, Richmond. Div. of Litter & Recycling.

    This publication, the last in a series of seven for elementary schools, is an environmental education curriculum guide with a focus on waste management issues. It contains a unit of exercises selected for sixth grade students focusing on community solutions to solid waste pollution. Waste management activities included in this unit seek to…

  6. 324 Building liquid waste handling and removal system project plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-07-29

    This report evaluates the modification options for handling radiological liquid waste generated during decontamination and cleanout of the 324 Building. Recent discussions indicate that the Hanford site railroad system will be closed by the end of FY 1998 necessitating the need for an alternate transfer method. The issue of handling of Radioactive Liquid Waste (RLW) from the 324 Building (assuming the 340 Facility is not available to accept the RLW) has been examined in at least two earlier engineering studies (Parsons 1997a and Hobart 1997). Each study identified a similar preferred alternative that included modifying the 324 Building RLWS to allow load-out of wastewater to a truck tanker, while making maximum use of existing piping, tanks, instrumentation, controls and other features to minimize costs and physical changes to the building. This alternative is accepted as the basis for further discussion presented in this study. The goal of this engineering study is to verify the path forward presented in the previous studies and assure that the selected alternative satisfies the 324 Building deactivation goals and objectives as currently described in the project management plan. This study will also evaluate options available to implement the preferred alternative and select the preferred option for implementation of the entire system. Items requiring further examination will also be identified. Finally, the study will provide a conceptual design, schedule and cost estimate for the required modifications to the 324 Building to allow removal of RLW. Attachment 5 is an excerpt from the project baseline schedule found in the Project Management Plan.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Fang Dong

    1997-01-01

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

  10. Study of alternative methods for the management of liquid scintillation counting wastes

    International Nuclear Information System (INIS)

    Roche-Farmer, L.

    1980-02-01

    The Nuclear Engineering Waste Disposal Site in Richland, Washington, is the only radioactive waste disposal facility that will accept liquid scintillation counting wastes (LSCW) for disposal. That site is scheduled to discontinue receiving LSCW by the end of 1982. This document explores alternatives presently available for management of LSCW: evaporation, distillation, solidification, conversion, and combustion

  11. MECHANISMS GOVERNING TRANSIENTS FROM THE BATCH INCINERATION OF LIQUID WASTES IN ROTARY KILNS

    Science.gov (United States)

    When "containerized" liquid wastes, bound on sorbents. are introduced into a rotary kiln in a batch mode, transient phenomena in-volving heat transfer into, and waste mass transfer out of, the sorbent can oromote the raoid release of waste vaoor into the kiln environment. This ra...

  12. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    Science.gov (United States)

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  13. Liquid radioactive wastes from hospitals by polymeric membrane; Tratamiento de residuos liquidos radiactivos hospitalarios mediante membranas polimericas

    Energy Technology Data Exchange (ETDEWEB)

    Arnal, J M; Sancho, M; Verdu, G [Universidad Politecnica de Valencia (Spain); Campayo, J M [LAINSA (Spain)

    1998-12-01

    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)

  14. Cnoidal waves as solutions of the nonlinear liquid drop model

    International Nuclear Information System (INIS)

    Ludu, Andrei; Sandulescu, Aureliu; Greiner Walter

    1997-01-01

    By introducing in the hydrodynamic model, i.e. in the hydrodynamic equation and the corresponding boundary conditions, the higher order terms in the deviation of the shape, we obtain in the second order the Korteweg de Vries equations (KdV). The same equation is obtained by introducing in the liquid drop model (LDM), i.e. in the kinetic, surface and Coulomb terms, the higher terms in the second order. The KdV equation has the cnoidal waves as steady-state solutions. These waves could describe the small anharmonic vibrations of spherical nuclei up to the solitary waves. The solitons could describe the preformation of clusters on the nuclear surface. We apply this nonlinear liquid drop model to the alpha formation in heavy nuclei. We find an additional minimum in the total energy of such systems, corresponding to the solitons as clusters on the nuclear surface. By introducing the shell effects we choose this minimum to be degenerated with the ground state. The spectroscopic factor is given by ratio of the square amplitudes in the two minima. (authors)

  15. Removal of cesium using coconut fiber in raw and modified forms for the treatment of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Jesus, Nella N.M. de; Nobre, Vanessa B.; Potiens Junior, Ademar J.; Sakata, Solange K.; Di Vitta, Patricia B.

    2013-01-01

    Sorption is one of the most studied methods to reduce the volume of radioactive waste streams. Cesium-137 is a radioisotope formed by the fission of uranium and it can cause health problems due to its easy assimilation by cells. The aim of this study is to evaluate the potential of coconut fiber in removing cesium from radioactive liquid wastes; this process can help in disposing radioactive waste. The experiments were performed in batch and the particle size of the fiber ranged between 0.30 mm and 0.50 mm. The fiber was treated with hydrogen peroxide in alkaline medium. The following parameters were analyzed: contact time, pH and concentration of cesium ions in aqueous solution. After the experiments the samples were filtered and cesium remaining in solution was quantified by inductively coupled plasma optical emission spectrometry. (author)

  16. Removal of cesium using coconut fiber in raw and modified forms for the treatment of radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Jesus, Nella N.M. de; Nobre, Vanessa B.; Potiens Junior, Ademar J.; Sakata, Solange K., E-mail: sksakata@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Di Vitta, Patricia B. [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Quimica

    2013-07-01

    Sorption is one of the most studied methods to reduce the volume of radioactive waste streams. Cesium-137 is a radioisotope formed by the fission of uranium and it can cause health problems due to its easy assimilation by cells. The aim of this study is to evaluate the potential of coconut fiber in removing cesium from radioactive liquid wastes; this process can help in disposing radioactive waste. The experiments were performed in batch and the particle size of the fiber ranged between 0.30 mm and 0.50 mm. The fiber was treated with hydrogen peroxide in alkaline medium. The following parameters were analyzed: contact time, pH and concentration of cesium ions in aqueous solution. After the experiments the samples were filtered and cesium remaining in solution was quantified by inductively coupled plasma optical emission spectrometry. (author)

  17. Thermodynamics of dilute aqueous solutions of imidazolium based ionic liquids

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Tejwant [Salt and Marine Chemicals Division, Central Salt and Marine Chemicals Research Institute, Council of Scientific and Industrial Research (CSIR), G.B. Marg, Bhavnagar 364002 (India); Kumar, Arvind, E-mail: arvind@csmcri.or [Salt and Marine Chemicals Division, Central Salt and Marine Chemicals Research Institute, Council of Scientific and Industrial Research (CSIR), G.B. Marg, Bhavnagar 364002 (India)

    2011-06-15

    Research highlights: The thermodynamic behaviour of aqueous imidazolium ILs has been investigated. Volumetric and ultrasonic results indicated the hydrophobic hydration of ILs. Viscometric studies revealed studied ionic liquids as water-structure makers. Hydration number increased with increase in alkyl chain length of the cation. - Abstract: Experimental measurements of density {rho}, speed of sound u, and viscosity {eta} of aqueous solutions of various 1-alkyl-3-methylimidazolium based ionic liquid (IL) solutions have been performed in dilute concentration regime at 298.15 K to get insight into hydration behaviour of ILs. The investigated ILs are based on 1-alkyl-3-methylimidazolium cation, [C{sub n}mim] having [BF{sub 4}]{sup -}, [Cl]{sup -}, [C{sub 1}OSO{sub 3}]{sup -}, and [C{sub 8}OSO{sub 3}]{sup -} as anions where n = 4 or 8. Several thermodynamic parameters like apparent molar volume {phi}{sub V}, isentropic compressibility {beta}{sub s}, and viscosity B-coefficients have been derived from experimental data. Limiting value of apparent molar volume has been discussed in terms of intrinsic molar volume (V{sub int}) molar electrostriction volume (V{sub elec}), molar disordered (V{sub dis}), and cage volume (V{sub cage}). Viscosity B-coefficients have been used to quantify the kosmotropic or chaotropic nature of ILs. Hydration number of ILs obtained using elctrostriction volume, isentropic compressibility, viscosity, and differential scanning calorimetry have been found to be comparative within the experimental error. The hydrophobic hydration has found to play an important role in hydration of ILs as compared to hydration due to hydrogen bonding and electrostriction. Limiting molar properties, hydration numbers, and B-coefficients have been discussed in terms of alkyl chain length of cation or nature of anion.

  18. Thermodynamics of dilute aqueous solutions of imidazolium based ionic liquids

    International Nuclear Information System (INIS)

    Singh, Tejwant; Kumar, Arvind

    2011-01-01

    Research highlights: → The thermodynamic behaviour of aqueous imidazolium ILs has been investigated. → Volumetric and ultrasonic results indicated the hydrophobic hydration of ILs. → Viscometric studies revealed studied ionic liquids as water-structure makers. → Hydration number increased with increase in alkyl chain length of the cation. - Abstract: Experimental measurements of density ρ, speed of sound u, and viscosity η of aqueous solutions of various 1-alkyl-3-methylimidazolium based ionic liquid (IL) solutions have been performed in dilute concentration regime at 298.15 K to get insight into hydration behaviour of ILs. The investigated ILs are based on 1-alkyl-3-methylimidazolium cation, [C n mim] having [BF 4 ] - , [Cl] - , [C 1 OSO 3 ] - , and [C 8 OSO 3 ] - as anions where n = 4 or 8. Several thermodynamic parameters like apparent molar volume φ V , isentropic compressibility β s , and viscosity B-coefficients have been derived from experimental data. Limiting value of apparent molar volume has been discussed in terms of intrinsic molar volume (V int ) molar electrostriction volume (V elec ), molar disordered (V dis ), and cage volume (V cage ). Viscosity B-coefficients have been used to quantify the kosmotropic or chaotropic nature of ILs. Hydration number of ILs obtained using elctrostriction volume, isentropic compressibility, viscosity, and differential scanning calorimetry have been found to be comparative within the experimental error. The hydrophobic hydration has found to play an important role in hydration of ILs as compared to hydration due to hydrogen bonding and electrostriction. Limiting molar properties, hydration numbers, and B-coefficients have been discussed in terms of alkyl chain length of cation or nature of anion.

  19. Electrochemical ion-exchange for active liquid waste treatment

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  20. Low-level liquid waste decontamination by ion exchange

    International Nuclear Information System (INIS)

    Campbell, D.O.; Lee, D.D.; Dillow, T.A.

    1991-12-01

    Improved processes are being developed to treat contaminated liquid wastes that have been and continue to be generated at Oak Ridge National Laboratory. Both inorganic and organic ion-exchange methods have given promising results. Nickel and cobalt hexacyanoferrate(2) compounds are extremely selective for cesium removal, with distribution coefficients in excess of 10 6 and remarkable insensitivity to competition from sodium and potassium. They tend to lose effectiveness at pH > ∼11, but some formulations are useful for limited periods of time up to pH ∼13. Sodium titanate is selective for strontium removal at high pH. The separations are so efficient that simple batch processes can yield large decontamination factors while generating small volumes of solid waste. A resorcinol-based resin developed at the Savannah River Site gave superior cesium removal, compared with other organic ion exchangers; the distribution coefficient was limited primarily by competition from potassium and was nearly independent of sodium. The optimum pH was ∼12.5. It was much less effective for strontium removal, which was limited by competition from sodium. 8 refs., 6 figs., 9 tabs

  1. Packed bed reactor treatment of liquid hazardous and mixed wastes

    International Nuclear Information System (INIS)

    Tennant, R.A.; Wantuck, P.J.; Vargas, R.

    1992-01-01

    We are developing thermal-based packed bed reactor (PBR) technology as an alternative to incineration for treatment of hazardous organic liquid wastes. The waste streams targeted by this technology are machining fluids contaminated with chlorocarbons and/or chlorofluorocarbons and low levels of plutonium or tritium The PBR offers several distinct advantages including simplistic design, rugged construction, ambient pressure processing, economical operations, as well as ease of scalability and maintainability. In this paper, we provide a description of the apparatus as well as test results using prepared mixtures of machining oils/emulsions with trichloroethylene (TCE), carbon tetrachloride (CCl 4 ), trichloroethane (TCA), and Freon TF. The current treatment system is configured as a two stage device with the PBR (1st stage) coupled to a silent discharge plasma (SDP) cell. The SDP serves as a second stage for further treatment of the gaseous effluent from the PBR. One of the primary advantages of this two stage system is that its suitability for closed loop operation where radioactive components are well contained and even CO 2 is not released to the environment

  2. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions.

    Science.gov (United States)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-01

    With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent's concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I2/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5M I2/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe4BTBP showed good removal of mercury, with an extraction efficiency of 97.5 ± 0.7% being achieved in a single stage. Better removal of mercury was achieved in a single stage using the extractants Cyanex 302 and Cyanex 923 in kerosene, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  3. Characterization of actinide-bearing sediments underlying liquid waste disposal facilities at Hanford

    International Nuclear Information System (INIS)

    Price, S.M.; Ames, L.L.

    1975-09-01

    Past liquid waste disposal practices at the U. S. Energy Research and Development Administration's Hanford Reservation have included the discharges of solutions containing trace quantities of actinides directly into the ground via structures collectively termed ''trenches''. Characterization of samples from two of these trenches, the 216-Z-9 and the 216-Z-1A(a), has been initiated to determine the present form and migration potential of plutonium stored in sediments which received high salt, acidic waste liquids. Analysis of samples acquired by drilling has revealed that the greatest measured concentration of Pu, approximately 10 6 μCi 239 Pu/liter of sediment, occurs in both facilities just below the points of release of the waste liquids. This concentration decreases to approximately 10 3 μCi 239 Pu/liter of sediment within the first 2 meters of the underlying sediment columns and to approximately 10 μCi 239 Pu/liter of sediment at the maximum depth sampled (9 meters). Examination of relatively undisturbed sediment cores illustrated two types of Pu occurrence responsible for this distribution. One of these types is composed of Pu particles (greater than 70 wt percent PuO 2 ) added to the disposal site in the same form. This ''particulate'' type was ''filtered out'' within the upper 1 meter of the sediment column, accounting for the high concentration of Pu/liter of sediment in this region. The second type of Pu (less than 0.5 wt percent PuO 2 ) was originally disposed of as soluble Pu(IV). This ''nonparticulate'' type penetrated deeper within the sediment profile and was deposited in association with silicate hydrolysis of the sediment fragments

  4. Distribution of the active liquid waste discharge concentration

    International Nuclear Information System (INIS)

    Chan, A.H.C.

    1985-03-01

    In assessing the proposal for removing the on-line liquid effluent monitor (LEM) from the Darlington NGS-A design, it was required to estimate the probability that the concentration of β-y emitters in the active liquid waste (ALW) tank discharges exceeds a specified level. To achieve this, it was necessary to know the underlying distribution of the ALW discharge concentration. Since the distribution could only be estimated from the historical data, it was also important to provide the confidence interval for the estimated probability. Using the ALW discharge records of Pickering and Bruce NGS-A, it was found that the log-normal distribution provided the best fit for the data. The frequency of the tank concentration exceeding the specified level of 24000μCi/m 3 was estimated to be 1 in 200,000 years at Bruce NGS-A and 1 in 100,000 years at Pickering. The 99% upper confidence limits are 1 in 2777 years and 1 in 77 years, respectively

  5. Biodegradation of bituminous products from processing liquid radioactive wastes

    International Nuclear Information System (INIS)

    Tibensky, L.; Krejci, F.; Hladky, E.; Halama, D.

    1988-01-01

    One of the possible ways of disturbing the stability of bituminous products from liquid radioactive waste processing, is biodegradation caused by common microorganisms. Pseudomonas bacteria and a Bacillus cereus culture were selected for experimental study of cultivation of microorganisms. Experiments with mixed cultures were also performed. Pitches, ajatin and imidazoline were used as inhibitors. The thin layer and the emulsion methods were used in assessing biological corrosion. The results of the experiments are discussed with respect to the dependence of bacterial growth on bitumen biodegradation, the effect of pH on bitumen degradation and the effect of inhibitors on bitumen biodegradation. The salts contained in bituminous products were not found to significantly affect the rate of destruction. The degree of degradation was found to mainly depend on the bitumen, its chemical composition, and on the conditions of storage. It was also found that inhibitor additions in some cases modified the properties of the matrix such that it became more liquid. The coefficient of extractibility thus increased of matrix salts. The recultivation of bacteria on a full-value medium resulted in the loss of the inhibitory effect. In some cases, the inhibitor even stimulated the growth of microorganisms. The use of inhibitors in an effort to achieve biostability of bituminous products thus did not solve the problem. (Z.M.). 2 tabs., 9 refs

  6. Microwave Plasma Enhanced Chemical Vapor Deposition of Diamond in Vapor of Methanol-Based Liquid Solutions

    National Research Council Canada - National Science Library

    Tzeng, Yonhua

    2000-01-01

    .... Liquid solutions are prepared by mixing methanol with other carbon containing liquid compounds which contain a greater than one ratio of carbon to oxygen such as acetone, ethanol, and iso-propanol...

  7. 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.)

  8. Colloid Genesis/Transport and Flow Pathway Alterations Resulting From Interactions of Reactive Waste Solutions and Hanford Vadose Zone Sediments

    International Nuclear Information System (INIS)

    Wan, Jiamin; Tokunaga, Tetsu K.

    2001-01-01

    Leakage of underground tanks containing high-level nuclear waste solutions has been identified at various DOE facilities. The Hanford Site is one the main facilities of concern, with about 2,300 to 3,400 m3 of leaked waste liquids. Radionuclides and other contaminants have been found in elevated concentrations in the vadose zone and groundwater underneath single shell tank farms. We do not currently know the mechanisms responsible for the unexpected deep migration of some contaminants through the vadose zone, and such understanding is urgently needed for planning remediation. Due to the extreme chemical conditions of the tank waste solutions (very high pH, aluminum concentration, and ionic strength), interactions between the highly reactive waste solutions and sediments underneath the tanks can result in dissolution of primary minerals of the sediments and precipitation of secondary phases including colloidal particles. Contaminants can sorb onto and/or co-precipitate with the secondary phases. Therefore transport of strongly associated contaminants on mobile colloids can be substantially greater than without colloids. The overall objective of this research is to improve our understanding on the effects of interactions between the tank waste solution and sediments on deep contaminant migration under Hanford Site conditions. This objective will be achieved through the following four tasks: (1) colloid generation and transport studies, (2) studies on sediment permeability and chemical composition alterations, (3) quantifying associations of contaminants with secondary colloids, and (4) studies on the combined effects of the aforementioned processes on deep contaminant migration

  9. New sorption-reagent materials for decontamination of liquid radioactive waste

    International Nuclear Information System (INIS)

    Avramenko, V.A.; Golikov, A.P.; Zheleznov, V.V.; Kaplun, E.V.; Marinin, D.V.; Sokolnitskaya, T.A.

    2001-01-01

    Full text: Use of selective sorbents in liquid radioactive waste (LRW) management is widely spread in the field of nuclear power objects liquid waste decontamination, since the main objective there is to remove long-lived radionuclides of the nuclear cycle. The latter include, first of all, cesium-137, strontium-90, cobalt-60 and a number of α-irradiators. In this case LRW composition for most of the nuclear power objects is rather simple, except acidic deactivation solutions. At the same time, liquid radioactive wastes of different research centers have a variable chemical and radiochemical composition depending on objectives and tasks of a given center research activities. As a result, application of sorption technologies in such waste decontamination determines special requirements to these sorbents selectivity: a wide spectrum of radionuclides that can be removed and fairly high selectivity enabling to remove radionuclides from solutions of complex chemical composition (containing surfactants, complexing agents etc.). This paper is concerned with studying properties of new materials selective to different radionuclides. These materials are capable to interact with solution components whether already contained in the waste or deliberately added into resulting solution. Such sorption-reagent materials combine universal character of co-precipitation methods with simplicity of sorption methods. In this work we studied sorption-reagent inorganic ion-exchange materials interacting with sulfate-, carbonate-, oxalate-, sulfide-, and permanganate-ions. Insoluble compounds formed as a result of this interaction increase tens- and hundreds-fold the sorption selectivity of different radionuclides - strontium, cobalt, mercury, iron, and manganese as compared to conventional ion-exchange system. By means of X-ray phase analysis, IR-spectroscopy, chemical and radiochemical analysis, we have studied the mechanism of radionuclide sorption on different sorption

  10. 30 CFR 250.217 - What solid and liquid wastes and discharges information and cooling water intake information must...

    Science.gov (United States)

    2010-07-01

    ... What solid and liquid wastes and discharges information and cooling water intake information must accompany the EP? The following solid and liquid wastes and discharges information and cooling water intake... 30 Mineral Resources 2 2010-07-01 2010-07-01 false What solid and liquid wastes and discharges...

  11. Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1984-01-01

    The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

  12. Review of Potential Candidate Stabilization Technologies for Liquid and Solid Secondary Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; Mattigod, Shas V.; Westsik, Joseph H.; Serne, R. Jeffrey; Icenhower, Jonathan P.; Scheele, Randall D.; Um, Wooyong; Qafoku, Nikolla

    2010-01-30

    Pacific Northwest National Laboratory has initiated a waste form testing program to support the long-term durability evaluation of a waste form for secondary wastes generated from the treatment and immobilization of Hanford radioactive tank wastes. The purpose of the work discussed in this report is to identify candidate stabilization technologies and getters that have the potential to successfully treat the secondary waste stream liquid effluent, mainly from off-gas scrubbers and spent solids, produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Down-selection to the most promising stabilization processes/waste forms is needed to support the design of a solidification treatment unit (STU) to be added to the Effluent Treatment Facility (ETF). To support key decision processes, an initial screening of the secondary liquid waste forms must be completed by February 2010.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-15

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Yeotikar, R.G.

    2007-01-01

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

  16. 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

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

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1996-07-01

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

  18. Using copper hexacyanoferrate (II) impregnated zeolite for cesium removal from radioactive liquid waste

    International Nuclear Information System (INIS)

    Fumio, K.; Kenji, M.

    1982-01-01

    Experiments were performed to obtain fundamental data on cesium ion removal characteristics of metal hexacyanoferrate (II) impregnated zeolite in radioactive liquid waste containing a large amount of sodium sulfate. Copper hexacyanoferrate (II) impregnated zeolite (CuFZ) was prepared and showed a high selectivity for cesium ion. The material was suitable for use in an ion exchange column. This exchanger could selectively and efficiently remove the cesium even if there is 15 wt% Na 2 SO 4 in the solution. Cesium removal ability and stability of CuFZ were excellent over a wide pH range between 1.5 and 10. The cesium ion exchange ability was not influenced by the presence of the alkali metal ions, calcium and magnesium, and carbonate ions even at concentrations 25 times greater than the cesium ion. However, since ammonium ion behaves similarly to cesium ion and interrupts latter ion adsorption, the presence of ammonium ion is not desirable. The CuFZ offers the possibility of separating and removing cesium from liquid wastes produced in facilities handling radioactive materials

  19. Development studies for the treatment of ORNL low-level liquid waste

    International Nuclear Information System (INIS)

    Campbell, D.O.; Lee, D.D.; Dillow, T.A.

    1991-11-01

    An experimental program is under way to investigate potential separation methods for application to specific problems relating to the management of low-level liquid wastes (LLLWs) at ORNL. This report summarizes experimental results that were acquired during fiscal year 1990 and have not been previously reported elsewhere. Measurements are presented for cesium and strontium removal from simulated high-salt waste compositions, using both inorganic ion- exchange sorbents and organic ion-exchange resins, and for one experiment with actual LLLW supernate solution from Melton Valley Storage Tank W-26, using inorganic sorbents. The purpose of the study was to acquire an extensive data base to support the development of flowsheets for decontamination of the LLLW currently stored at ORNL. Experimental measurements with inorganic ion exchangers focused on batch separations of cesium using several transition-metal hexacyanoferrate(2) compositions (ferrocyanides) and of strontium using titanium oxide-based sorbents. Cesium distribution coefficients in the range of 1 x 10 6 were generally observed with nickel and cobalt ferrocyanides at pH values ≤11, yielding DFs of about 100 with 100 ppm sorbent in a single-stage batch separation. Most organic ion-exchange resins are not very effective for cesium removal from such high salt concentrations, but a new resorcinol-based resin developed at the Savannah River Site was found to be considerably superior to any other such material tested. Several chelating resins were effective for removing strontium from the waste simulants. An ion-exchange column test successfully demonstrated the simultaneous removal of both cesium and strontium from a waste simulant solution

  20. Use of complexones solutions in liquid carbon dioxide for cleaning of materials contaminated with heavy and radioactive metals

    International Nuclear Information System (INIS)

    Shadrin, A.Yu.; Kamachev, V.A.; Kiseleva, R.N.; Murzin, A.A.; Shafikov, D.N.; Bondin, V.V.; Efremov, I.V.; Kovalev, D.N.; Podoinitsyn, S.V.

    2003-01-01

    I n this paper liquid carbon dioxide (pressure 50-70 atm) was used for decontamination. The performed experiments on removal of cobalt, nickel, uranium and americium nitrates and carbonates by different solutions have shown that the solutions of such complexing agents as hexafluoroacetylacetone (HFA), tributylphosphate (TBP), di-2-ethylhexylphosphoric acid (D2EHPA) in liquid CO 2 can be used for purification of pulps, metals, paper and fabrics. Liquid CO 2 is high viscosity of the medium and hence low diffusion coefficients and long duration of the processes. It is known that 20 minutes are sufficient to attain equilibrium in supercritical CO 2 medium on metal removal by HFA solutions. During the experiments it was established that with the use of liquid CO 2 the keeping time should be increased to 40 min, which is acceptable from the standpoint of technical feasibility of decontamination processes in these solutions. Experiments on really contaminated samples of pulps, metals and fabrics have confirmed that the decontamination coefficients of 30-100 can be easily obtained by 2-3 fold material treatment operations. The secondary waste volume therewith is less by a factor of 20-200 than that of traditional techniques. (authors)

  1. The Influence of Extractant TOA, Stirring Time on the Extraction ProcessLiquid-liquid, and Liquid Membrane on the Liquid Wastes Containing Cd

    International Nuclear Information System (INIS)

    Prayitno; Djoko-Sardjono; Nurimaniwati; Adhe-Helmayani

    2000-01-01

    The influence of extractant and stirring time on the reduction componentcadmium on liquid wastes has been investigated. The method of experimentalused the extraction with liquid membrane emulsion. The parameters to beinvestigated were extractant amount tri-n octylamine (TOA), duration ofstirring time. In this investigated, extractant amount was varied from 5 to25 % (v/v) TOA, duration of stirring time varied from 5 to minutes. Theresult of experimental can be concluded that the best condition obtained forreducing cadmium component was on extractant amount 20 % (v/v) TOA, stirringtime 25 minutes. The best condition for reducing the cadmium component wasefficiency factor 98.35%. (author)

  2. (Liquid + liquid) phase equilibrium and critical behavior of binary solution {heavy water + 2,6-dimethylpyridine}

    International Nuclear Information System (INIS)

    Xu, Chen; Chai, Shouning; Yin, Tianxiang; Chen, Zhiyun; Shen, Weiguo

    2015-01-01

    Highlights: • Coexistence curves, heat capacities and turbidities were measured. • Deuterium effect on coexistence curves was discussed. • Universal critical amplitude ratios were tested. • Asymmetry of coexistence curves was analyzed by the complete scaling theory. - Abstract: The (liquid + liquid) coexistence curves, the isobaric heat capacities per unit volume and the turbidities for the binary solution of {heavy water + 2,6-dimethylpyridine} have been precisely measured. The values of the critical exponents were obtained, which confirmed the 3D-Ising universality. It was found that the critical temperature dropped by 5.9 K and the critical amplitude of the coexistence curve significantly increased as compared to the binary solution of {water + 2,6-dimethylpyridine}. The complete scaling theory was applied to well describe the asymmetric behavior of the diameter of the coexistence curve as the heat capacity contribution was considered. Moreover, the values of the critical amplitudes of the correlation length and the osmotic compressibility were deduced, which together with the critical amplitudes of the coexistence curve and the heat capacity to test universal amplitude ratios

  3. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    International Nuclear Information System (INIS)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-01-01

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I 2 /KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I 2 /KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe 4 BTBP showed good removal of mercury, with an

  4. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    Energy Technology Data Exchange (ETDEWEB)

    Tunsu, Cristian, E-mail: tunsu@chalmers.se; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-15

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I{sub 2}/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I{sub 2}/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe{sub 4}BTBP showed good removal of mercury

  5. 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.).

  6. Innovative waste management solutions: An outlook for the future

    Energy Technology Data Exchange (ETDEWEB)

    1993-01-01

    At a conference on various aspects of waste management, papers were presented on the effects of landfills, plastic debris in the marine environment, skills development and training in the waste industry, composting, remediation of contaminated soils, disposal methods, sludge derived byproducts, recycling, waste management economics, municipal solid waste management, ship-generated waste, managing waste in national parks, and septic tank sludge treatment. Separate abstracts have been prepared for five papers from this conference.

  7. Determination of free acid in high level liquid wastes by means of fixed pH value

    International Nuclear Information System (INIS)

    Li Jifu; Duan Shirong; Wu Xi; Yu Xueren

    1991-01-01

    For the determination of free acid in high level liquid wastes, 8% potassium oxalate solution with pH 6.50 as a complex agent of hydrolizable ion is added to 1 AW and the solution is titrated with standard sodium hydroxide to reach the original pH value. The quantity of free acid is calculated by standard sodium hydroxide consumed. This method is simple, rapid and accurate. The relative error of analysis is less than ±4%. The average percentage of recovery is 99.6-101.0%

  8. Recovery of fission products from waste solutions utilizing controlled cathodic potential electrolysis

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.

    1975-01-01

    Fission products, e.g., palladium, rhodium and technetium, are recovered from aqueous waste solutions thereof, e.g., aged Purex alkaline waste solutions. The metal values from the waste solutions are extracted by ion exchange techniques. The metals adsorbed by the ion exchange resin are eluted and selectively recovered by controlled cathodic potential electrolysis. The metal values deposited on the cathode are recovered and, if desired, further purified

  9. Magnetic precipitate separation for Ni plating waste liquid using HTS bulk magnets

    Energy Technology Data Exchange (ETDEWEB)

    Oka, T., E-mail: okat@eng.niigata-u.ac.jp [Niigata University, 8050 Ikarashi-Ninocho, Nishi-ku, Niigata 950-2181 (Japan); Kimura, T.; Mimura, D.; Fukazawa, H.; Fukui, S.; Ogawa, J.; Sato, T.; Ooizumi, M. [Niigata University, 8050 Ikarashi-Ninocho, Nishi-ku, Niigata 950-2181 (Japan); Yokoyama, K. [Ashikaga Institute of Technology, 268-1 Ohmae-cho, Ashikaga, Tochigi 326-8558 (Japan); Tsujimura, M. [Aichi Giken Co., 2-1-47 Shiobaru, Minami-ku, Fukuoka 815-8520 (Japan); Terasawa, T. [IMRA Material R and D Co., Ltd., 2-1 Asahimachi, Kariya, Aichi 448-0032 (Japan)

    2013-01-15

    Highlights: ► The magnetic separation was operated for recycling the electroless plating waste. ► The HTS bulk magnet effectively attracted the ferromagnetic precipitates with Ni. ► The separation ratios over 90% were reported under flow rates up to 1.35 L/min. -- Abstract: The magnetic separation experiment for recycling the nickel-bearing precipitates in the waste liquid from the electroless plating processes has been practically conducted under the high gradient magnetic separation technique with use of the face-to-face HTS bulk magnet system. A couple of facing magnetic poles containing Sm123 bulk superconductors were activated through the pulsed field magnetization process to 1.86 T at 38 K and 2.00 T at 37 K, respectively. The weakly magnetized metallic precipitates of Ni crystals and Ni–P compounds deposited from the waste solution after heating it and pH controlling. The high gradient magnetic separation technique was employed with the separation channels filled with the stainless steel balls with dimension of 1 and 3 mm in diameter, which periodically moved between and out of the facing magnetic poles. The Ni-bearing precipitates were effectively attracted to the magnetized ferromagnetic balls. We have succeeded in obtaining the separation ratios over 90% under the flow rates less than 1.35 L/min.

  10. Radiation exposure rate and liquid level measurement inside a high level liquid waste (HLLW) storage tank

    International Nuclear Information System (INIS)

    Sur, B.; Yue, S.; Thekkevarriam, A.

    2007-01-01

    An instrument based on an inexpensive, small silicon diode has been developed and used to measure, for the first time, the gamma radiation exposure rate profile inside a 6.4 mm diameter reentrant thermo-well tube, immersed in the highly radioactive liquid solution in an HLLW storage tank. The measurement agrees with previous calculations of exposure rate, and provides confirmation for safe and effective radiation work plans and material selection for investigations and remediation of the storage tank facility. The measured radiation exposure rate profile is also used to confirm that the position of tank internal structures have not changed because of aging and corrosion, and to obtain, within a few mm, the level of liquid inside the tank. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

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

  12. 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

  13. New sorbents and ion exchangers for nuclear waste solution remediation

    International Nuclear Information System (INIS)

    Clearfield, A.; Peng, G.Z.; Cahill, R.A.; Bellinghausen, P.; Aly, H.I.; Scott, K.; Wang, J.D.

    1993-01-01

    There is now a concerted effort underway to clean up the accumulated nuclear wastes as the major sites around the country. Because of the complexity of the mixtures in the holding tanks highly specific exchangers are required to fulfill a multitude of desired tasks. These include removal of Cs + , Sr 2+ , Tc, Actinides and possible recovery of rare and precious metals. No one exchanger or sequestrant can accomplish these tasks and a variety of exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers and new organo-inorganic exchangers towards Cs + , Sr 2+ and rare-earth ions in acid and basic media will be described. Preliminary data on the effect of high levels of sodium nitrate on the uptake of these ions will also be presented, as well as the changes observed in selectivity in simulated waste solutions. A possible separation scheme based on these data will be described

  14. Ammonia nitrogen removal from aqueous solution by local agricultural wastes

    Science.gov (United States)

    Azreen, I.; Lija, Y.; Zahrim, A. Y.

    2017-06-01

    Excess ammonia nitrogen in the waterways causes serious distortion to environment such as eutrophication and toxicity to aquatic organisms. Ammonia nitrogen removal from synthetic solution was investigated by using 40 local agricultural wastes as potential low cost adsorbent. Some of the adsorbent were able to remove ammonia nitrogen with adsorption capacity ranging from 0.58 mg/g to 3.58 mg/g. The highest adsorption capacity was recorded by Langsat peels with 3.58 mg/g followed by Jackfruit seeds and Moringa peels with 3.37 mg/g and 2.64 mg/g respectively. This experimental results show that the agricultural wastes can be utilized as biosorbent for ammonia nitrogen removal. The effect of initial ammonia nitrogen concentration, pH and stirring rate on the adsorption process were studied in batch experiment. The adsorption capacity reached maximum value at pH 7 with initial concentration of 500 mg/L and the removal rate decreased as stirring rate was applied.

  15. Solid on liquid deposition, a review of technological solutions

    OpenAIRE

    Homsy, Alexandra; Laux, Edith; Jeandupeux, Laure; Charmet, Jérôme; Bitterli, Roland; Botta, Chiara; Rebetez, Yves; Banakh, Oksana; Keppner, Herbert

    2015-01-01

    Solid-on-liquid deposition (SOLID) techniques are of great interest to the MEMS and NEMS (Micro- and Nano Electro Mechanical Systems) community because of potential applications in biomedical engineering, on-chip liquid trapping, tunable micro-lenses, and replacements of gate oxides. However, depositing solids on liquid with subsequent hermetic sealing is difficult because liquids tend to have a lower density than solids. Furthermore, current systems seen in nature lack thermal, mechanical or...

  16. Reuse of waste beer yeast sludge for biosorptive decolorization of reactive blue 49 from aqueous solution.

    Science.gov (United States)

    Wang, Baoe; Guo, Xiu

    2011-06-01

    Reactive blue 49 was removed from aqueous solution by biosorption using powder waste sludge composed of Saccharomyces cerevisiae from the beer-brewing industry. The effect of initial pH, temperature and the biosorption thermodynamics, equilibrium, kinetics was investigated in this study. It was found that the biosorption capacity was at maximum at initial pH 3, that the effect of temperature on biosorption of reactive blue 49 was only slight in relation to the large biosorption capacity (25°C, 361 mg g(-1)) according as the biosorption capacity decreased only 43 mg g(-1) at the temperature increased from 25 to 50°C. The biosorption was spontaneous, exothermic in nature and the dye molecules movements decreased slightly in random at the solid/liquid interface during the biosorption of dye on biosorbents. The biosorption equilibrium data could be described by Freundich isotherm model. The biosorption rates were found to be consistent with a pseudo-second-order kinetics model. The functional group interaction analysis between waste beer yeast sludge and reactive blue 49 by the aid of Fourier transform infrared (abbr. FTIR) spectroscopy indicated that amino components involved in protein participated in the biosorption process, which may be achieved by the mutual electrostatic adsorption process between the positively charged amino groups in waste beer yeast sludge with negatively charged sulfonic groups in reactive blue 49.

  17. The Addition of Hatchery Liquid Waste to Dairy Manure Improves Anaerobic Digestion

    Directory of Open Access Journals (Sweden)

    WRT Lopes

    Full Text Available ABSTRACT The objective of this study was to determine the optimal inclusion level of liquid egg hatchery waste for the anaerobic co-digestion of dairy cattle manure. A completely randomized experimental was applied, with seven treatments (liquid hatchery waste to cattle manure ratios of0: 100, 5:95, 10:90, 15:85, 20:80, 25:75 and 30:70, with five replicates (batch digester model each. The evaluated variables were disappearance of total solids (TS, volatile solids (VS, and neutral detergent fiber (NDF, and specific production of biogas and of methane. Maximum TS and VS disappearance of 41.3% and 49.6%, were obtained at 15.5% and 16.0% liquid hatchery waste inclusion levels. The addition of 22.3% liquid hatchery considerably reduced NDF substrate content (53.2%. Maximum specific biogas production was obtained with 17% liquid hatchery waste, with the addition of 181.7 and 229.5 L kg-1TS and VS, respectively. The highest methane production, at 120.1 and 151.8 L CH4 kg-1TS and VS, was obtained with the inclusion of 17.5 and 18.0% liquid hatchery waste, respectively. The addition of liquid hatchery waste atratios of up to 15.5%in co-digestion with cattle manure reduced solid and fiber levels in the effluent, and improved biogas and methane production.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Ion exchanger material based on Titanium phosphate for liquid radioactive waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Maslova, M.; Gerasimova, L. [Tananaev Institute of Chemistry and Technology of Rare Elements and Mineral Resources, Kola Scientific Center, Russian Academy of Sciences, Apatity (Russian Federation)

    2013-07-01

    A comparative study of the physicochemical and service properties of samples of Ti(OH) 1.36(HPO4) 1.32 * 2.3H2O sorbent in the finely dispersed and granulated forms, mastered for commercial production, was made. The sorption of Cs and Sr cations from solutions of various, compositions was studied in batch experiments, and the diffusion coefficients of the exchanging ions were determined. The hydrolytic stability of the. sorbents was examined with the aim to determine the optimal operation conditions. Experiments showed that the cation exchangers based on titanium phosphate are the most efficient in removal from liquid radioactive waste of induced radioactive isotopes of corrosion products, which is due to formation of weakly dissociating compounds of nonferrous metal ions with functional groups of the ion exchangers in the sorbent phase. (author)