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.)

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)

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)

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

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

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

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)

Natural diatomite process for removal of radioactivity from liquid waste

Energy Technology Data Exchange (ETDEWEB)

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

2007-01-15

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

Natural diatomite process for removal of radioactivity from liquid waste

International Nuclear Information System (INIS)

Osmanlioglu, Ahmet Erdal

2007-01-01

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

Natural diatomite process for removal of radioactivity from liquid waste.

Science.gov (United States)

Osmanlioglu, Ahmet Erdal

2007-01-01

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

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.

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

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.

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)

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

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)

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)

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

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.

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

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 )

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.

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.)

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

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

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.)

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

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)

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.)

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.

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

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)

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.

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

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

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.

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

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

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)

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)

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)

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.)

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.)

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.)

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.

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.

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)

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

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)

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

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.)

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

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)

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

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

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

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)

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

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

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.

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

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

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

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

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)

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.)

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)

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.)

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 K_d (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.

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.

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.

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.)

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

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

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

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

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

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.

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

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)

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

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)

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

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

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.)

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

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.

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

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.

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

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)

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.

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.

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

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

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

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.)

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.)

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

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.)

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

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.

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.

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.

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

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

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.

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.

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.

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...

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.

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

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

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

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

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

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.

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

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

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)

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.)

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)

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

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.

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.)

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.)

Management of hot cell waste in Atalante Facilities (abstract and presentation slides)

International Nuclear Information System (INIS)

Dancausse, Jean-Philippe; Ferlay, Gilles; Eysseric, Catherine

2005-01-01

In solution R and D experiments on nuclear fuel from dissolution to liquid extraction lead to produce a large set of wastes. This paper present how these highly contaminated solid and liquid wastes is managed in Hot Cells and in Atalante. Firstly, an inventory of several types of generated wastes is made: 1) Solid wastes. 2) Glass reactors and liquid solution containers. 3) Plastic and Teflon materials for sampling, Highly corrosive solutions. 4) Metallic containers for solid storage like fuels, crucibles. 5) Miscellaneous mixed solid materials. 6) Liquid wastes. 7) Rinsing liquids. 8) Highly corrosive waste containing fluorhydric acid. 9) Analytical solution with sulphate ions. 10) Organic solvent coming from liquid-liquid extraction. A focus will be made on optimised treatment of 1) solid wastes: Mechanically and chemically 2) liquid wastes containing sulphate ions and hydrogen fluoride, 3) organic liquid waste: to remove activity before hydrothermal oxidation. (Author)

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

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

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

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

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.

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

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

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

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

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

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

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

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.

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

Supported liquid membrane based removal of lead(II) and cadmium(II) from mixed feed: Conversion to solid waste by precipitation

Energy Technology Data Exchange (ETDEWEB)

Bhatluri, Kamal Kumar; Manna, Mriganka Sekhar; Ghoshal, Aloke Kumar; Saha, Prabirkumar, E-mail: p.saha@iitg.ac.in

2015-12-15

Highlights: • Simultaneous removal of two heavy metals lead and cadmium. • Conversion of liquid waste to solid precipitation. • Precipitation facilitates the metals transportation through LM. • Solidification of liquid waste minimizes the final removal of waste. - Abstract: Simultaneous removal of two heavy metals, lead(II) and cadmium(II), from mixed feed using supported liquid membrane (SLM) based technique is investigated in this work. The carrier-solvent combination of “sodium salt of Di-2-ethylhexylphosphoric acid (D2EHPA) (4% w/w) in environmentally benign coconut oil” was immobilized into the pores of solid polymeric polyvinylidene fluoride (PVDF) support. Sodium carbonate (Na{sub 2}CO{sub 3}) was used as the stripping agent. Carbonate salts of lead(II) and cadmium(II) were formed in the stripping side interface and they were insoluble in water leading to precipitation inside the stripping solution. The transportation of solute is positively affected due to the precipitation. Lead(II) removal was found to be preferential due to its favorable electronic configuration. The conversion of the liquid waste to the solid one was added advantage for the final removal of hazardous heavy metals.

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.)

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

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

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)

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.)

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.)

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

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.

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

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

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

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.)

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

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

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)

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

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)

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.

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.

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

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

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.

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

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.

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.)

Recovery of actinides from TBP-Na2Co3 scrub-waste solutions: the ARALEX process

International Nuclear Information System (INIS)

Horwitz, E.P.; Bloomquist, C.A.A.; Mason, G.W.; Leonard, R.A.; Ziegler, A.A.

1979-08-01

A flowsheet for the recovery of actinides from TBP-Na 2 CO 3 scrub-waste solutions has been developed, based on batch extraction data, and tested, using laboratory-scale countercurrent extraction techniques. The process, called the ARALEX process, uses 2-ethyl-1-hexanol (2-EHOH) to extract the TBP degradation products (HDBP and H 2 MBP) from acidified Na 2 CO 3 scrub waste leaving the actinides in the aqueous phase. Dibutyl and monobutyl phosphoric acids are attached to the 2-EHOH molecules through hydrogen bonds, which also diminish the ability of the HDBP and H 2 MBP to complex actinides. Thus all actinides remain in the aqueous raffinate. Dilute sodium hydroxide solutions can be used to back-extract the dibutyl and monobutyl phosphoric acid esters as their sodium salts. The 2-EHOH can then be recycled. After extraction of the acidified carbonate waste with 2-EHOH, the actinides may be readily extracted from the raffinate with DHDECMP or, in the case of tetra- and hexavalent actinides, with TBP. The ARALEX process can also be applied to other actinide waste streams which contain appreciable concentrations of polar organic compounds (e.g., detergents) that interfere with conventional actinide ion exchange and liquid-liquid extraction procedures. 20 figures, 6 tables

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.)

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)

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...

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)

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)

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)

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.)

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

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

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

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...

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)

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.

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)

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

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

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.

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 ...

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

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

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

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

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

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

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

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

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)

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)

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.

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.

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Processing method for liquid waste containing various kinds of radioactive material

International Nuclear Information System (INIS)

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

1996-01-01

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

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.

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

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

Acid digestion of organic liquids

International Nuclear Information System (INIS)

Partridge, J.A.; Bosuego, G.P.

1980-10-01

Laboratory studies on the destruction of liquid organic wastes by acid digestion are discussed. A variety of liquid waste types was tested, including those encountered in the nuclear industry as well as some organic liquids representative of non-nuclear industrial wastes. The liquids tested were vacuum pump oil, tri-n-butyl phosphate (TBP), normal paraffin hydrocarbon solvent (NPH), a mixture of 30 vol% TBP in NPH, carbon tetrachloride (CCl 4 ), trichloroethane, toluene, hexone (methyl isobutyl ketone), a mixture of hexone and NPH, polychlorobiphenyl (PCB), isopropanol, normal-decane, and two waste organic solutions from Hanford radioactive waste tanks. The tests demonstrated that several types of organic liquids can be destroyed by the acid digestion process. 8 figures, 19 tables

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

International Nuclear Information System (INIS)

Wakaida, Yasuo.

1996-01-01

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

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

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.)

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, Wilbur O.

1985-01-01

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

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, W.O.

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

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)

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

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.

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.

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

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.)

Physico-chemical treatment of liquid waste on an industrial plant for electrocoagulation.

Science.gov (United States)

Mlakar, Matej; Levstek, Marjetka; Stražar, Marjeta

2017-10-01

Wastewater from washing, oil separators, the metal processing and detergent industries, was tested and treated for treatment of different types of liquid waste at industrial level at Domžale-Kamnik Wastewater Treatment Plant (WWTP). The effect of implementing the electrocoagulation (EC) and flotation processes, respectively, is analysed and includes the duration of the EC implementation, voltage, number of electrodes, and chemical addition, as well as the pH effect and conductivity. The tests were performed not only on various types of liquid waste, but also on different mixtures of liquid waste. Laboratory analysis of the samples before and after EC have shown an effective reduction not only in organic loads in accordance with the COD (chemical oxygen demand) parameter, but also in mineral oil content, toxic metal concentration, and surfactants. The COD in liquid waste from the detergent industry was reduced by 73% and the content of surfactants by 64%. In liquid waste from the metal processing industry, the COD decreased by up to 95%, while the content of toxic metals decreased from 59 to 99%. Similar phenomena were shown in liquid waste from oil separators, where the COD was reduced to 33% and the concentration of mineral oils by 99%. Some of the liquid wastes were mixed together in the ratio 1:1, thus allowing testing of the operation of EC technology in heterogeneous liquid waste, where the final result proved to be effective cleaning as well. After treatment in the process of EC, the limit values of the treated water proved appropriate for discharge into the sewerage system.

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)

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

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.

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

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

Development of a test system for high level liquid waste partitioning

Directory of Open Access Journals (Sweden)

Duan Wu H.

2015-01-01

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

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)

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

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)

Chemical decontamination method for radioactive metal waste

International Nuclear Information System (INIS)

Onuma, Tsutomu; Tanaka, Akio; Shibuya, Sadao.

1991-01-01

When contaminants mainly composed of copper remained on the surface of stainless steel wastes sent from an electrolytic reduction as a first step are chemically decontaminated, metal wastes are discriminated to carbon steel wastes and stainless steel wastes. Then, the carbon steel wastes are applied only with the first step of immersing in a sulfuric acid solution, and stainless steel wastes are applied with a first step of immersing into a sulfuric acid solution for electrolytic reduction for a predetermined period of time and a second step of immersing into a liquid in which an oxidative metal salt is added to sulfuric acid. The decontamination liquid which is used for immersing the stainless steel wastes in the second step and the oxidation force of which is lowered is used as the sulfuric acid solution in the first step for the carbon steel wastes. In view of the above, the decontamination liquid of the second step can be utilized most effectively, enabling to greatly decrease the secondary wastes and to improve decontamination efficiency. (T.M.)

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.)

Filtration of Oak Ridge National Laboratory simulated liquid low-level waste

International Nuclear Information System (INIS)

Fowler, V.L.; Hewitt, J.D.

1989-08-01

A method for disposal of Oak Ridge National Laboratory's (ORNL's) liquid low-level radioactive waste (LLLW) is being developed in which the material will be solidified in cement and stored in an aboveground engineered storage facility. The acceptability of the final waste form rests in part on the presence or absence of transuranic isotopes. Filtration methods to remove transuranic isotopes from the bulk liquid stored in the Melton Valley Storage Tanks (MVST) were investigated in this study. Initial batch studies using waste from MVST indicate that >99.9% of the transuranic isotopes can be removed from the bulk liquid by simple filtration. Bench-scale studies with a nonradioactive surrogate waste indicate that >99.5% of the suspended solids can be removed from the bulk liquid via inertial crossflow filtration. 4 refs., 3 figs., 11 tabs

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.)

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.)

Stabilization and isolation of low-level liquid waste disposal sites

International Nuclear Information System (INIS)

Phillips, S.J.; Gilbert, T.W.

1987-01-01

Rockwell Hanford Operations is developing and testing equipment for stabilization and isolation of low-level radioactive liquid waste disposal sites. Stabilization and isolation are accomplished by a dynamic consolidation and particulate grout injection system. System equipment components include: a mobile grout plant for transport, mixing, and pumping of particulate grout; a vibratory hammer/extractor for consolidation of waste, backfill, and for emplacement of the injector; dynamic consolidation/injector probe for introducing grout into fill material; and an open-void surface injector that uses surface or subsurface mechanical or pneumatic packers and displacement gas filtration for introducing grout into disposal structure access piping. Treatment of a liquid-waste disposal site yields a physically stable, cementitious monolith. Additional testing and modification of this equipment for other applications to liquid waste disposal sites is in progress

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.)

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

Development of Characterization Protocol for Mixed Liquid Radioactive Waste Classification

International Nuclear Information System (INIS)

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

2017-01-01

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

Development of characterization protocol for mixed liquid radioactive waste classification

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-29

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

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.).

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.)

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)

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.

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.

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

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.

Decontamination factor Improvement and Waste Reduction of Full-scaled Evaporation System for Liquid Radioactive Waste Treatment

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki Tae; Ju, Young Jong; Seol, Jeung Gun; Cho, Nam Chan [KNF, Daejeon (Korea, Republic of); Ha, Dong Hwan; Kim, Yun Kwan [Jeontech Co., Suwon (Korea, Republic of)

2016-05-15

Liquid radioactive waste is produced from nuclear power plants, nuclear research centers, radiopharmaceuticals and nuclear fuel fabrication plants, etc. Ion-exchange, chemical precipitation, evaporation, filtration, liquid/solid extraction and centrifugal are applied to treat the liquid waste. Chemical precipitation requires low capital and operation cost. However, it produces large amount of secondary waste and has low DF (decontamination factor). Evaporation process removes variety of radionuclides in high DF. But, it also has problems in scaling and foaming [3, 4]. In this study, it is investigated that the effect of switching lime precipitation and centrifugal processes to evaporation system for improvement of removal efficiency and decrease of waste in full-scaled radioactive wastewater treatment plant. By swapping full-scaled wastewater treatment system from the centrifugal and the lime precipitation to the evaporator and the crystallizer in the nuclear fuel fabrication plant, it was possible to increase removal efficiency and to minimize waste productivity. Radioactivity concentration of effluent is decreased from 0.01 Bq/mL to ND level. Besides, waste production was reduced from 15 drums/yr to 2 drums/yr (87%).

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

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)

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)

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.)

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)

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)

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.)

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)

Stabilization of liquid low-level and mixed wastes: a treatability study

International Nuclear Information System (INIS)

Carson, S.; Cheng, Yu-Cheng; Yellowhorse, L.; Peterson, P.

1996-01-01

A treatability study has been conducted on liquid low-level and mixed wastes using the stabilization agents Aquaset, Aquaset II, Aquaset II-H, Petroset, Petroset-H, and Petroset and Petroset II. A total of 40 different waste types with activities ranging from 10 -14 to 10 -4 curies/ml have been stabilized. Reported data for each waste include its chemical and radiological composition and the optimum composition or range of compositions (weight of agent/volume of waste) for each stabilization agent used. All wastes were successfully stabilized with one or more of the stabilization agents and all final waste forms passed the Paint Filter Liquids Test (EPA Method 9095)

Isolation of Metals from Liquid Wastes: Reactive in Turbulent Thermal Reactors

International Nuclear Information System (INIS)

Wendt, Jost O.L.

2001-01-01

A Generic Technology for treatment of DOE Metal-Bearing Liquid Waste The DOE metal-bearing liquid waste inventory is large and diverse, both with respect to the metals (heavy metals, transuranics, radionuclides) themselves, and the nature of the other species (annions, organics, etc.) present. Separation and concentration of metals is of interest from the standpoint of reducing the volume of waste that will require special treatment or isolation, as well as, potentially, from the standpoint of returning some materials to commerce by recycling. The variety of metal-bearing liquid waste in the DOE complex is so great that it is unlikely that any one process (or class of processes) will be suitable for all material. However, processes capable of dealing with a wide variety of wastes will have major advantages in terms of process development, capital, and operating costs, as well as in environmental and safety permitting. Moreover, to the extent that a process operates well with a variety of metal-bearing liquid feedwastes, its performance is likely to be relatively robust with respect to the inevitable composition variations in each waste feed. One such class of processes involves high-temperature treatment of atomized liquid waste to promote reactive capture of volatile metallic species on collectible particulate substrates injected downstream of a flame zone. Compared to low-temperature processes that remove metals from the original liquid phase by extraction, precipitation, ion exchange, etc., some of the attractive features of high-temperature reactive scavenging are: The organic constituents of some metal-bearing liquid wastes (in particular, some low-level mixed wastes) must be treated thermally in order to meet the requirements of the Resource Conservation and Recovery Act (RCRA) and Toxic Substances Control Act (TSCA), and the laws of various states. No species need be added to an already complex liquid system. This is especially important in light of the fact

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

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

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)

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.

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

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

U.S. Department of Energy's initiatives for proliferation prevention in Russia: results of radioactive liquid waste treatment project, year 2

International Nuclear Information System (INIS)

Pokhitonov, Y.; Kamachev, V.; Kelley, D.

2010-10-01

The objective of the project is to engage weapons scientists with training and research programs at selected nuclear sites in Russia and apply high technology polymers to immobilize legacy ILW and HLW liquids that have posed environmental challenges over the years. One compelling advantage of the projects is that V.G. Khlopin Radium Institute and Pacific Nuclear Solutions have been engaged in applied research for seven years to validate the performance and effectiveness of the polymer technology for use with radioactive liquids. With conclusive results of the research work on sixty active and simulant waste streams, the project can focus on actual applications of the technology at Ozersk (Mayak), Sever sk (SCC) Zheleznogorsk (MCC) and Gatchyna rather that on pure research. The long term objective of the project is find viable waste management solutions for serious radioactive and chemical contamination that has existed in Russia and the U. S. for several decades. The polymer technologies may be applied to all radioactive liquid. This paper summarizes the experimental work of the immobilization process and data definition of the most effective polymer compositions in addition to determining the optimum polymer to liquid ratios for economic considerations. (Author)

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

A process for treatment of mixed waste containing chemical plating wastes

International Nuclear Information System (INIS)

Anast, K.R.; Dziewinski, J.; Lussiez, G.

1995-01-01

The Waste Treatment and Minimization Group at Los Alamos National Laboratory has designed and will be constructing a transportable treatment system to treat low-level radioactive mixed waste generated during plating operations. The chemical and plating waste treatment system is composed of two modules with six submodules, which can be trucked to user sites to treat a wide variety of aqueous waste solutions. The process is designed to remove the hazardous components from the waste stream, generating chemically benign, disposable liquids and solids with low level radioactivity. The chemical and plating waste treatment system is designed as a multifunctional process capable of treating several different types of wastes. At this time, the unit has been the designated treatment process for these wastes: Destruction of free cyanide and metal-cyanide complexes from spent plating solutions; destruction of ammonia in solution from spent plating solutions; reduction of Cr VI to Cr III from spent plating solutions, precipitation, solids separation, and immobilization; heavy metal precipitation from spent plating solutions, solids separation, and immobilization, and acid or base neutralization from unspecified solutions

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.)

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 ...

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

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.)

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

Energy Technology Data Exchange (ETDEWEB)

Vieira, Ludmila Cabreira

2016-07-01

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

Designing testing service at baristand industri Medan’s liquid waste laboratory

Science.gov (United States)

Kusumawaty, Dewi; Napitupulu, Humala L.; Sembiring, Meilita T.

2018-03-01

Baristand Industri 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 Industri Medan is liquid waste testing service. The company set the standard of service is nine working days for testing services. At 2015, 89.66% on testing services liquid waste does not meet the specified standard of services company because of many samples accumulated. The purpose of this research is designing online services to schedule the coming the liquid waste sample. The method used is designing an information system that consists of model design, output design, input design, database design and technology design. The results of designing information system of testing liquid waste online consist of three pages are pages to the customer, the recipient samples and laboratory. From the simulation results with scheduled samples, then the standard services a minimum of nine working days can be reached.

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

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.

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.

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.

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.

Functions and requirements document, WESF decoupling project, low-level liquid waste system

Energy Technology Data Exchange (ETDEWEB)

Rasmussen, J.H., Fluor Daniel Hanford

1997-02-27

The Waste Encapsulation and Storage Facility (WESF) was constructed in 1974 to encapsulate and store cesium and strontium which were isolated at B Plant from underground storage tank waste. The WESF, Building 225-B, is attached physically to the west end of B Plant, Building 221-B, 200 East area. The WESF currently utilizes B Plant facilities for disposing liquid and solid waste streams. With the deactivation of B Plant, the WESF Decoupling Project will provide replacement systems allowing WESF to continue operations independently from B Plant. Four major systems have been identified to be replaced by the WESF Decoupling Project, including the following: Low Level Liquid Waste System, Solid Waste Handling System, Liquid Effluent Control System, and Deionized Water System.

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

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.

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

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)

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

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

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.

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...

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

Application Of A Thin Film Evaporator System For Management Of Liquid High-Level Wastes At Hanford

International Nuclear Information System (INIS)

Tedeschi, A.R.; Wilson, R.A.

2010-01-01

A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD

Energy Technology Data Exchange (ETDEWEB)

TEDESCHI AR; WILSON RA

2010-01-14

A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy & Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

Solidification/stabilisation of liquid oil waste in metakaolin-based geopolymer

Energy Technology Data Exchange (ETDEWEB)

Cantarel, V.; Nouaille, F.; Rooses, A.; Lambertin, D., E-mail: david.lambertin@cea.fr; Poulesquen, A.; Frizon, F.

2015-09-15

Highlights: • Formulation with 20 vol.% of oil in a geopolymer have been successful tested. • Oil waste is encapsulated as oil droplets in metakaolin-based geopolymer. • Oil/geopolymer composite present good mechanical performance. • Carbon lixiviation of oil/geopolymer composite is very low. - Abstract: The solidification/stabilisation of liquid oil waste in metakaolin based geopolymer was studied in the present work. The process consists of obtaining a stabilised emulsion of oil in a water-glass solution and then adding metakaolin to engage the setting of a geopolymer block with an oil emulsion stabilised in the material. Geopolymer/oil composites have been made with various oil fraction (7, 14 and 20 vol.%). The rigidity and the good mechanical properties have been demonstrated with compressive strength tests. Leaching tests evidenced the release of oil from the composite material is very limited whereas the constitutive components of the geopolymer (Na, Si and OH{sup −}) are involved into diffusion process.

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%

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.)

Technetium recovery from high alkaline solution

Energy Technology Data Exchange (ETDEWEB)

Nash, Charles A.

2016-07-12

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

Radioactive waste management in a fuel reprocessing facility in fiscal 1982

International Nuclear Information System (INIS)

1984-01-01

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

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

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)

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)

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

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)

Management of radioactive liquid waste at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Bendixsen, C.L.

1992-01-01

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

Membrane preparation and process development for radioactive waste treatment

Energy Technology Data Exchange (ETDEWEB)

Lee, K. W.; Kim, G. W.; Kim, S. K. [KAERI, Daejeon (Korea, Republic of); and others

2012-01-15

The membrane manufacturing technology with hydrophilic function that can minimize fouling when it applies to the radioactive liquid waste treatment process was developed. Thermodynamic and rheological analysis for polysulfone casting solution containing polyvinylpyrrolidone was performed. On the basis of the results of preparation of the hydrophilic polymer membrane solution, the hollow fiber membrane for radioactive liquid waste treatment was manufactured and its performance analysis was carried out. As a results, it turns out the hydrophilic hollow fiber membrane has more 90 % of flux increment effect and also more 2.5 times fouling reducing effect than one prepared with only polysulfone. In addition, as investigating the separation property of radioactive liquid waste for the electrofilteration membrane process, a proper range for application of radioactive liquid wastes was established through the thorough electrofiltration analysis of various wastes containing metal salt, surfactants and oil.

Membrane preparation and process development for radioactive waste treatment

International Nuclear Information System (INIS)

Lee, K. W.; Kim, G. W.; Kim, S. K.

2012-01-01

The membrane manufacturing technology with hydrophilic function that can minimize fouling when it applies to the radioactive liquid waste treatment process was developed. Thermodynamic and rheological analysis for polysulfone casting solution containing polyvinylpyrrolidone was performed. On the basis of the results of preparation of the hydrophilic polymer membrane solution, the hollow fiber membrane for radioactive liquid waste treatment was manufactured and its performance analysis was carried out. As a results, it turns out the hydrophilic hollow fiber membrane has more 90 % of flux increment effect and also more 2.5 times fouling reducing effect than one prepared with only polysulfone. In addition, as investigating the separation property of radioactive liquid waste for the electrofilteration membrane process, a proper range for application of radioactive liquid wastes was established through the thorough electrofiltration analysis of various wastes containing metal salt, surfactants and oil

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)

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)

Liquid radwaste treatment by microfiltration, ultrafiltration and reverse osmosis

International Nuclear Information System (INIS)

Dulama, M.; Deneanu, N.; Popescu, I.V.

2001-01-01

Radioactive liquid waste processing is an integral part of any facility involved in nuclear power generation, radioisotope production, research and development, decontamination or other aspects of nuclear energy. The aqueous liquid radwastes from the decontamination center are currently treated by the membrane plant. Generally, the liquid waste streams are effectively volume-reduced by a combination of continuous crossflow microfiltration (MF), spiral wound reverse osmosis (SWRO) and tubular reverse osmosis membrane technologies. Backwash chemical cleaning wastes from the membrane plant are further volume-reduced by evaporation. The concentrate from the membrane plant is ultimately immobilized with bitumen. We performed experiments using two simulated waste solution; secondary waste from the decontamination process with POD (Permanganate Oxidation Decontamination) solution and secondary waste from decontamination with CAN-DECON solution. The experimental tests have been done with cellulose acetate (CA) membrane and polysulfonate (PSF) membrane manufactured at Research Center for Macromolecular Materials and Membranes Bucharest and with Millipore membrane type VS 0.025 μm. A schematic of the laboratory-scale test facility is presented

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)

Distribution of aquifers, liquid-waste impoundments, and municipal water-supply sources, Massachusetts

Science.gov (United States)

Delaney, David F.; Maevsky, Anthony

1980-01-01

Impoundments of liquid waste are potential sources of ground-water contamination in Massachusetts. The map report, at a scale of 1 inch equals 4 miles, shows the idstribution of aquifers and the locations of municipal water-supply sources and known liquid-waste impoundments. Ground water, an important source of municipal water supply, is produced from shallow sand and gravel aquifers that are generally unconfined, less than 200 feet thick, and yield less than 2,000 gallons per minute to individual wells. These aquifers commonly occupy lowlands and stream valleys and are most extensive in eastern Massachusetts. Surface impoundments of liquid waste are commonly located over these aquifers. These impoundments may leak and allow waste to infiltrate underlying aquifers and alter their water quality. (USGS)

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. 

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.

A mobile system for treating low-salinity low-activity liquid wastes

International Nuclear Information System (INIS)

Sobolev, I.A.; Timofeev, E.M.; Panteleev, V.I.; Karlin Yu.V.; Kropotov, V.N.; Slastennikov, Yu.T.; Chuikov, V.Yu.; Demkin, V.I.; Rozhkov, V.T.

1993-01-01

Radioactive wastes are produced not only in radiochemical production and nuclear power stations but also in numerous research institutes and industrial organizations. The specific activities of these wastes are low, and the volumes do not exceed a few dozen cubic meters a year at each individual organization, but processing such territorially distributed wastes is complicated. This particularly applies to liquid wastes, whose transportation involves a high risk of contamination if the sealing fails. As a rule, liquid wastes are solidified before transportation to a storage site. In some cases, that simplified approach leads to an unduly large consumption of solidifying materials, and particularly to an increase in volume, while storage is an expensive technique. A considerable volume reduction in the wastes to be stored is provided by processing the liquid wastes to concentrate the radionuclides in a small volume, with the main volume of treated water discharged to the drains. Two styles are possible: a stationary plant for processing wastes at each institution or a mobile one with a centralized service base, e.g., at the storage site. Mobile systems have been reported in world practice, although there is no detailed information on them. From the economic viewpoint, the second approach is preferable because it enables one to conduct such operations with fewer plants and fewer staff. That a mobile concept that was used at the Moscow Radon Cooperative in 1985 in processing liquid wastes at regional storage locations is summarized in this article. Research and development led in 1989 to the manufacture of a prototype mobile system mounted on an MAZ articulated vehicle, which included three basic modules: ultrafiltration, electrodialysis, and filtration ones. Each module is located on a separate framework and is connected to the others by reinforced rubber hoses

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

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

Devoluming method of acidic radioactive liquid waste and processing system therefor

International Nuclear Information System (INIS)

Shirai, Takamori; Honda, Tadahiro

1998-01-01

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

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

Directory of Open Access Journals (Sweden)

Putri Wening Ratrinia

2017-02-01

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

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

Directory of Open Access Journals (Sweden)

Putri Wening Ratrinia

2016-12-01

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

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)

Method for stabilizing low-level mixed wastes at room temperature

Science.gov (United States)

Wagh, Arun S.; Singh, Dileep

1997-01-01

A method to stabilize solid and liquid waste at room temperature is provided comprising combining solid waste with a starter oxide to obtain a powder, contacting the powder with an acid solution to create a slurry, said acid solution containing the liquid waste, shaping the now-mixed slurry into a predetermined form, and allowing the now-formed slurry to set. The invention also provides for a method to encapsulate and stabilize waste containing cesium comprising combining the waste with Zr(OH).sub.4 to create a solid-phase mixture, mixing phosphoric acid with the solid-phase mixture to create a slurry, subjecting the slurry to pressure; and allowing the now pressurized slurry to set. Lastly, the invention provides for a method to stabilize liquid waste, comprising supplying a powder containing magnesium, sodium and phosphate in predetermined proportions, mixing said powder with the liquid waste, such as tritium, and allowing the resulting slurry to set.

Solidification of liquid concentrate and solid waste generated as by-products of the liquid radwaste treatment systems in light-water reactors

International Nuclear Information System (INIS)

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

1977-01-01

The treatment of liquid concentrate and solid waste produced in light-water reactors as by-products of liquid radwaste treatment systems consists of five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging (solidification) and waste package handling. This paper will concern itself primarily with the solidification operation, however, the other operations enumerated as well as the types of wastes treated and their origins will be briefly described, especially with regards to their effects on solidification. During solidification, liquid concentrate and solid wastes are incorporated with a solidification agent to form a monolithic, free-standing solid. The basic solidification agent types either currently used in the United States or proposed for use include absorbants, hydraulic cement, urea-formaldehyde, other polymer systems, and bitumen. The operation, formulations and limitations of these agents as used for radwaste solidification will be discussed. Properties relevant to the evaluation of solidified waste forms will be identified and relative comparisons made for wastes solidified by various processes

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

2008-01-01

We studied cement-like solidification process for low-level liquid waste with relatively high levels of radioactivity 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 of the planned repository, and this leads to cracking during repository storage. It is also preferable that the pH of the pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Therefore, the present 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. In the FLS procedure, silica fume and blast furnace slag are added to the liquid waste containing BaSO 4 precipitate. We show the range of appropriate Ba/SO 4 molar ratio is from 1.1 to 1.5 in the present solidification process by leaching tests for some kinds of solid waste samples. The CIS reaction yield is over 98% at a typical CIS condition, i.e. Ba/SO 4 molar ratio=1.3, reaction temperature=60 deg C, and time=3 hr. (author)

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

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

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

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

Process for removing sulfate anions from waste water

Science.gov (United States)

Nilsen, David N.; Galvan, Gloria J.; Hundley, Gary L.; Wright, John B.

1997-01-01

A liquid emulsion membrane process for removing sulfate anions from waste water is disclosed. The liquid emulsion membrane process includes the steps of: (a) providing a liquid emulsion formed from an aqueous strip solution and an organic phase that contains an extractant capable of removing sulfate anions from waste water; (b) dispersing the liquid emulsion in globule form into a quantity of waste water containing sulfate anions to allow the organic phase in each globule of the emulsion to extract and absorb sulfate anions from the waste water and (c) separating the emulsion including its organic phase and absorbed sulfate anions from the waste water to provide waste water containing substantially no sulfate anions.

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

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

Oak Ridge National Lebroatory Liquid&Gaseous Waste Treatment System Strategic Plan

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, S.D.

2003-09-09

Excellence in Laboratory operations is one of the three key goals of the Oak Ridge National Laboratory (ORNL) Agenda. That goal will be met through comprehensive upgrades of facilities and operational approaches over the next few years. Many of ORNL's physical facilities, including the liquid and gaseous waste collection and treatment systems, are quite old, and are reaching the end of their safe operating life. The condition of research facilities and supporting infrastructure, including the waste handling facilities, is a key environmental, safety and health (ES&H) concern. The existing infrastructure will add considerably to the overhead costs of research due to increased maintenance and operating costs as these facilities continue to age. The Liquid Gaseous Waste Treatment System (LGWTS) Reengineering Project is a UT-Battelle, LLC (UT-B) Operations Improvement Program (OIP) project that was undertaken to develop a plan for upgrading the ORNL liquid and gaseous waste systems to support ORNL's research mission.

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.

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.

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

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)

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)

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...

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

International Nuclear Information System (INIS)

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

2009-01-01

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

The fixation of radioactive wastes in cement

International Nuclear Information System (INIS)

Kulichenko, V.V.; Dukhovich, F.S.; Volkova, O.I.; Boyarinova, M.V.

1976-01-01

The authors study the leaching behaviour of the main long-lived fission products 90 Sr and 137 Cs. It is found that 90 Sr and 137 Cs have high elution values, namely (2-12) x 10 -2 resp. (2-6) x 10 -2 g/cm 2 /24h, independently of the type of waste. On the basis of these results, maximum concentrations for the solutions in the cement/solution mixtures are proposed. Further studies relate to the formation of radiolysis gas in the waste fixed to cement. Experiments are described to make use of the empty space in the containers, filled with solid waste by filling them with mixtures of cement and liquid radioactive waste of 10 -4 to 1- 6 Ci. The ratio solution/cement should amount to 0.5. The containers are then buried underground. This method of combined waste storage helped to reduce the cost for the storage of liquid waste by about 40-50%. (RB) [de

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

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

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

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)

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Air Emissions Sampling from Vacuum Thermal Desorption for Mixed Wastes Designated with a Combustion Treatment Code for the Energy Solutions LLC Mixed Waste Facility

International Nuclear Information System (INIS)

Christensen, M.E.; Willoughby, O.H.

2009-01-01

EnergySolutions LLC is permitted by the State of Utah to treat organically-contaminated Mixed Waste by a vacuum thermal desorption (VTD) treatment process at its Clive, Utah treatment, storage, and disposal facility. The VTD process separates organics from organically-contaminated waste by heating the material in an inert atmosphere, and captures them as concentrated liquid by condensation. The majority of the radioactive materials present in the feed to the VTD are retained with the treated solids; the recovered aqueous and organic condensates are not radioactive. This is generally true when the radioactivity is present in solid form such as inorganic salts, metals or metallic oxides. The exception is when volatile radioactive materials are present such as radon gas, tritium, or carbon-14 organic chemicals. Volatile radioactive materials are a small fraction of the feed material. On August 28, 2006, EnergySolutions submitted a request to the USEPA for a variance to the Land Disposal Restrictions (LDR) standards for wastes designated with the combustion treatment code (CMBST). The final rule granting a site specific treatment variance was effective June 13, 2008. This variance is an alternative treatment standard to treatment by CMBST required for these wastes under USEPA's rules. The State of Utah provides oversight of the VTD processing operations. A demonstration test for treating CMBST-coded wastes was performed on April 29, 2008 through May 1, 2008. Three separate process cycles were conducted during this test. Both solid/liquid samples and emission samples were collected each day during the demonstration test. To adequately challenge the unit, feed material was spiked with trichloroethylene, o-cresol, dibenzofuran, and coal tar. Emission testing was conducted by EnergySolutions' emissions test contractor and sampling for radioactivity within the off-gas was completed by EnergySolutions' Health Physics department. This report discusses the emission testing

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