Deployment of wireless sensor network in pyrochemical processing of metallic fuels

International Nuclear Information System (INIS)

Baghyalakshmi, D.; Shrikrishnan, T.S.; Ebenezer, Jemimah; Madhusoodanan, K.; Satya Murty, S.A.V.; Vannia Perumal, S.; Venkatesh, P.; Prabhakara Reddy, B.

2016-01-01

With the advent of wireless sensor networking technology, industries started adapting the wireless monitoring systems in phases to measure and control various process parameters. To test the feasibility for implementing Wireless Sensor Network to measure the potentials of an electrochemical cell and the temperatures of actinide drawdown process at Pyrochemical process studies laboratory, at Chemistry Group, IGCAR, Kalpakkam, experiments have been carried out. An experimental setup with two Wireless Sensor Networking nodes has been deployed inside argon atmosphere glove boxes. The Electrorefining studies on U and U based alloys and the studies on actinide recovery from the electrolyte salt in actinide drawdown process are carried out in the glove box. The WSN measuring system was tested and validated by measuring the potentials of an electrochemical cell and the temperatures of actinide drawdown process. The WSN system is proposed to be installed in the hot cells of the Chemistry Group where irradiated U-Zr fuel is reprocessed. This paper briefs the need for remote measuring in pyrochemical reprocessing and validation of the remote signals by measuring the potentials of an electrochemical cell and the temperatures of the actinide draw down process. (author)

Aqueous recovery of plutonium from pyrochemical processing residues

International Nuclear Information System (INIS)

Gray, L.W.; Gray, J.H.

1984-01-01

Pyrochemical processes provide rapid methods to reclaim plutonium from scrap residues. Frequently, however, these processes yield an impure plutonium product and waste residues that are contaminated with actinides and are therefore nondiscardable. The Savannah River Laboratory and Plant and the Rocky Flats Plant are jointly developing new processes using both pyrochemistry and aqueous chemistry to generate pure product and discardable waste. An example of residue being treated is that from the molten salt extraction (MSE), a mixture of NaCl, KCl, MgCl 2 , PuCl 3 , AmCl 3 , PuO 2 , and Pu 0 . This mixture is scrubbed with molten aluminum containing a small amount of magnesium to produce a nonhomogeneous Al-Pu-Am-Mg alloy. This process, which rejects most of the NaCl-KCl-MgCl 2 salts, results in a product easily dissolved in 6M HNO 3 -0.1M HF. Any residual chloride in the product is removed by precipitation with Hg(I) followed by centrifuging. Plutonium and americium are then separated by the standard Purex process. The americium, initially diverted to the solvent extraction waste stream, can either be recovered or sent to waste

Molten salts processes and generic simulation

International Nuclear Information System (INIS)

Ogawa, Toru; Minato, Kazuo

2001-01-01

Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO 2 and PuO 2 in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

Molten salts processes and generic simulation

Energy Technology Data Exchange (ETDEWEB)

Ogawa, Toru; Minato, Kazuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO{sub 2} and PuO{sub 2} in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

Corrosion resistance of ceramic materials in pyrochemical reprocessing atmosphere by using molten salt for spent nuclear oxide fuel. Corrosion research under chlorine gas condition

International Nuclear Information System (INIS)

Takeuchi, Masayuki; Hanada, Keiji; Koizumi, Tsutomu; Aose, Shinichi; Kato, Toshihiro

2002-12-01

Pyrochemical reprocessing using molten salts (RIAR process) has been recently developed for spent nuclear oxide fuel and discussed in feasibility study. It is required to improve the corrosion resistance of equipments such as electrolyzer because the process is operated in severe corrosion environment. In this study, the corrosion resistance of ceramic materials was discussed through the thermodynamic calculation and corrosion test. The corrosion test was basically carried out in alkali molten salt under chlorine gas condition. And further consideration about the effects of oxygen, carbon and main fission product's chlorides were evaluated in molten salt. The result of thermodynamic calculation shows most of ceramic oxides have good chemical stability on chlorine, oxygen and uranyl chloride, however the standard Gibb's free energies with carbon have negative value. On the other hand, eleven kinds of ceramic materials were examined by corrosion test, then silicon nitride, mullite and cordierite have a good corrosion resistance less than 0.1 mm/y. Cracks were not observed on the materials and flexural strength did not reduce remarkably after 480 hours test in molten salt with Cl 2 -O 2 bubbling. In conclusion, these three ceramic materials are most applicable materials for the pyrochemical reprocessing process with chlorine gas condition. (author)

Pyrochemical processing of DOE spent nuclear fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1995-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

Development of pyrochemical spent fuel management in the UK

Energy Technology Data Exchange (ETDEWEB)

Banfield, Zara; Cogan, John; Farrant, Dave; Gaubert, Emmanuel; Hopkins, Phil; Lewin, Bob [BNFL - Nexia Solutions Limited, Workington Facility, B291 Trenches, Sellafield (United Kingdom)

2006-07-01

Nexia Solutions is undertaking a programme to investigate the role of pyrochemical techniques for spent nuclear fuel and legacy fuel management. The principal UK client is the energy unit, and the other clients are the Nuclear Decommissioning Authority (NDA), for legacy fuel conditioning, and BNFL's corporate investment in advanced reactor systems which is contributing to the Generation IV programme. The emphasis of the programme is pragmatic industrialization, which we have identified as key for the establishment of pyrochemical fuel management. From our experience operating fuel manufacture, power generation and reprocessing plant we know that the areas which require particular attention for successful implementation are: - Plant Interfaces, - Operability, - Process Definition, - Underpinning Science. Plant Interfaces encompass the definition of feeds, products, effluents and wastes and whether the process can meet the constraints they impose. Operability is concerned with the sustainability of the plant processes and is linked to the use of nil-maintenance continuous systems and elimination of batch / mechanical operations and maintenance. Process Definition focuses on the performance, control, recovery and safety of individual unit operations. Together these underpin industrial nuclear plant implementability. As an example, we have built a test rig to demonstrate molten salts transfers, since we consider this to be a capability without which pyrochemical processing will not be viable. Similarly, we have developed pilot scale electro-refiner designs for high continuous throughput and we are building development modules to underpin key features of the designs. Scientific work has been targeted at electro-refiner actinide partitioning and has been expanded to investigate other critical areas of the process which include efficient uranium / salt separation, salt clean up and the development of waste forms which perform at least as well as borosilicate glass

Development of pyrochemical spent fuel management in the UK

International Nuclear Information System (INIS)

Banfield, Zara; Cogan, John; Farrant, Dave; Gaubert, Emmanuel; Hopkins, Phil; Lewin, Bob

2006-01-01

Nexia Solutions is undertaking a programme to investigate the role of pyrochemical techniques for spent nuclear fuel and legacy fuel management. The principal UK client is the energy unit, and the other clients are the Nuclear Decommissioning Authority (NDA), for legacy fuel conditioning, and BNFL's corporate investment in advanced reactor systems which is contributing to the Generation IV programme. The emphasis of the programme is pragmatic industrialization, which we have identified as key for the establishment of pyrochemical fuel management. From our experience operating fuel manufacture, power generation and reprocessing plant we know that the areas which require particular attention for successful implementation are: - Plant Interfaces, - Operability, - Process Definition, - Underpinning Science. Plant Interfaces encompass the definition of feeds, products, effluents and wastes and whether the process can meet the constraints they impose. Operability is concerned with the sustainability of the plant processes and is linked to the use of nil-maintenance continuous systems and elimination of batch / mechanical operations and maintenance. Process Definition focuses on the performance, control, recovery and safety of individual unit operations. Together these underpin industrial nuclear plant implementability. As an example, we have built a test rig to demonstrate molten salts transfers, since we consider this to be a capability without which pyrochemical processing will not be viable. Similarly, we have developed pilot scale electro-refiner designs for high continuous throughput and we are building development modules to underpin key features of the designs. Scientific work has been targeted at electro-refiner actinide partitioning and has been expanded to investigate other critical areas of the process which include efficient uranium / salt separation, salt clean up and the development of waste forms which perform at least as well as borosilicate glass. Other

Immobilization of chloride-rich radioactive wastes produced by pyrochemical operations

International Nuclear Information System (INIS)

McDaniel, E.W.; Terry, J.W.

1997-08-01

A a result of its former role as a producer of nuclear weapons components, the Rocky Flats Environmental Technology Site (RFETS), Golden, Colorado accumulated a variety of plutonium-contaminated materials. When the level of contamination exceeded a predetermined level (the economic discard limit), the materials were classified as residues rather than waste and were stored for later recovery of the plutonium. Although large quantities of residues were processed, others, primarily those more difficult to process, remain in storage at the site. It is planned for the residues with lower concentrations of plutonium to be disposed of as wastes at an appropriate disposal facility, probably the Waste Isolation Pilot Plant (WIPP). Because the plutonium concentration is too high or because the physical or chemical form would be difficult to get into a form acceptable to WIPP, it may not be possible to dispose of a portion of the residues at WIPP. The pyrochemical salts are among the residues that are difficult to dispose of. For a large percentage of the pyrochemical salts, safeguards controls are required, but WIPP was not designed to accommodate safeguards controls. A potential solution would be to immobilize the salts. These immobilized salts would contain substantially higher plutonium concentrations than is currently permissible but would be suitable for disposal at WIPP. This document presents the results of a review of three immobilization technologies to determine if mature technologies exist that would be suitable to immobilize pyrochemical salts: cement-based stabilization, low-temperature vitrification, and polymer encapsulation. The authors recommend that flow sheets and life-cycle costs be developed for cement-based and low-temperature glass immobilization

Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

Energy Technology Data Exchange (ETDEWEB)

Steindler, M.J.; Ader, M.; Barletta, R.E.

1980-09-01

For pyrochemical and dry processing materials development included exposure to molten metal and salt of Mo-0.5% Ti-0.07% Ti-0.01% C, Mo-30% W, SiC, Si/sub 2/ON/sub 2/, ZrB/sub 2/-SiC, MgAl/sub 2/O/sub 4/, Al/sub 2/O/sub 3/, AlN, HfB/sub 2/, Y/sub 2/O/sub 3/, BeO, Si/sub 3/N/sub 4/, nickel nitrate-infiltrated W, W-coated Mo, and W-metallized alumina-yttria. Work on Th-U salt transport processing included solubility of Th in liquid Cd, defining the Cd-Th and Cd-Mg-Th phase diagrams, ThO/sub 2/ reduction experiments, and electrolysis of CaO in molten salt. Work on pyrochemical processes and associated hardware for coprocessing U and Pu in spent FBR fuels included a second-generation computer model of the transport process, turntable transport process design, work on the U-Cu-Mg system, and U and Pu distribution coefficients between molten salt and metal. Refractory metal vessels are being service-life tested. The chloride volatility processing of Th-based fuel was evaluated for its proliferation resistance, and a preliminary ternary phase diagram for the Zn-U-Pu system was computed. Material characterization and process analysis were conducted on the Exportable Pyrochemical process (Pyro-Civex process). Literature data on oxidation of fissile metals to oxides were reviewed. Work was done on chemical bases for the reprocessing of actinide oxides in molten salts. Flowsheets are being developed for the processing of fuel in molten tin. Work on encapsulation of solidified radioactive waste in metal matrix included studies of leach rate of crystalline waste materials and of the impact resistance of metal-matrix waste forms. In work on the transport properties of nuclear waste in geologic media, adsorption of Sr on oolitic limestone was studied, as well as the migration of Cs in basalt. Fitting of data on the adsorption of iodate by hematite to a mathematical model was attempted.

The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

International Nuclear Information System (INIS)

Ahluwalia, R.K.; Geyer, H.K.

1996-01-01

The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest

Materials and coating technology for pyrochemical reprocessing applications

International Nuclear Information System (INIS)

Jayakumar, T.; Kamachi Mudali, U.

2013-01-01

Metallic fuelled fast breeder reactors with co-located pyrochemical reprocessing plants have been proposed as the best option in order to increase the breeding gain, reduce the doubling time of the fuel and reprocess short cooled and high burnup fuel. To establish the pyrochemical reprocessing plants with various unit operations, it is necessary to identify, develop and qualify reliable corrosion resistant materials and coatings for service in molten LiCI-KCI salt and molten uranium environment operating at 773 to 1573 K. Towards materials and coating technology development and testing for molten salt environment a high temperature corrosion testing laboratory was established and studies were initiated. Molten salt test assembly for testing materials and coatings in molten LiCI-KCI salt under controlled ultra high pure (UHP) argon environment at high temperatures has been designed, fabricated, commissioned and tests were carried out on various candidate materials and coatings. Electro-formed (EF) Ni, Ni with Ni-W coating, coatings of ZrN, TiN, HfN and Ti-Si-N on high density (HD) graphite, candidate materials like 2.25Cr-1Mo steel, 9Cr-1Mo steel, 316L stainless steel, Ni base alloys (INCONEL 600, 625 and 690), HD graphite, pyrolytic graphite (PyG), and yttria stabilized zirconia (YSZ) and alumina-40wt% titania thermal barrier coatings were tested for their suitability for molten salt applications. Corrosion studies indicated that YSZ and PyG showed superior corrosion resistance in molten LiCI-KCI salt at 873 K up to 2000 h exposure. Surface modification techniques like annealing, laser remelting and laser shock processing were pursued to consolidate the coatings and improve their high temperature performance. Coating integrity using dielectric electrochemical system and thermal cycling furnace established that, compared to plain 9Cr-1Mo steel YSZ coated 9Cr-1Mo steel performed better from 473 K to 1223 K. The presentation highlights the results of the

Development of nitride fuel and pyrochemical process for transmutation of minor actinides

International Nuclear Information System (INIS)

Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo; Uno, Masayoshi

2010-01-01

Nitride fuel cycle for transmutation of minor actinides has been investigated under the double-strata fuel cycle concept. Mononitride solid solutions containing minor actinides have been prepared and characterised. Thermo-physical properties, such as thermal expansion, heat capacity and thermal diffusivity, have been measured by use of minor actinide nitride and burn-up simulated nitride samples. Irradiation behaviour of nitride fuel has been examined by irradiation tests. Pyrochemical process for treatment of spent nitride fuel has been investigated mainly by electrochemical measurements and nitride formation behaviour in pyrochemical process has been studied for recycled fuel fabrication. Recent results of experimental study on nitride fuel and pyrochemical process are summarised in the paper. (authors)

A Study on Pyrochemical Process for Treating Fuel Debris from the Fukushima-Daiichi Reactors

International Nuclear Information System (INIS)

Sakamura, Y.; Iizuka, M.; Koyama, T.; Kitawaki, S.; Nakayoshi, A.; Kofuji, H.

2015-01-01

After the severe accident at Fukushima-daiichi nuclear power plants, CRIEPI and JAEA started a feasibility study on the pyrochemical treatment of the corium that mainly consists of UO 2 -ZrO 2 solid solution. In this study, reduction behaviours of zirconium oxide compounds were investigated in LiCl-Li 2 O salt bath at 923 K. It was experimentally verified that uranium in the simulated corium could be reduced to the metallic form and a part of zirconium was converted to Li 2 ZrO 3 . At higher Li 2 O concentrations in LiCl, Li 2 ZrO 3 was converted to Li 6 Zr 2 O 7 and Li 8 ZrO 6 . In the subsequent electrorefining, Li 2 ZrO 3 reacts with UCl 3 dissolved in the electrolyte salt to give UO 2 precipitate. Therefore, how to remove the Li 2 ZrO 3 from the reduction product is a key point for the pyrochemical treatment of the corium. (authors)

Pyrochemical recovery of plutonium fluoride reduction slag

International Nuclear Information System (INIS)

Christensen, D.C.; Rayburn, J.A.

1983-07-01

A process was developed for the pyrochemical recovery of plutonium from residues resulting from the PuF 4 reduction process. The process involves crushing the CaF 2 slag and dissolving it at 800 0 C in a CaCl 2 solvent. The plutonium, which exists either as finely divided metal or as incompletely reduced fluoride salt, is reduced to metal and/or allowed to coalesce as a massive button in the bottom of the reaction crucible. The recovery of plutonium in a 1-day cycle averaged 96%; all of the resulting residues were discardable

Application of the pyrochemical DOS, developed by the CEA, within reprocessing of CERCER transmutation fuel targets

Energy Technology Data Exchange (ETDEWEB)

Mendes, E.; Ducasse, T.; Bertrand, M. [CEA, Centre de Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, SMCS, LDPS, F-30207 Bagnols-sur-Ceze (France); Miguirditchian, M. [CEA, Centre de Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, SMCS, LCPE, F-30207 Bagnols-sur-Ceze (France)

2016-07-01

Pyrochemical technology using high-temperature molten salts and molten metal media presents a potential interest for an overall separation and transmutation strategy for long-lived radionuclides. Within the frame of the two French acts on radioactive waste management, a pyrochemical research/development program was launched at the CEA Marcoule in the late 90's. The second step is the actinides back-extraction, which consists in a liquid/liquid oxidative stripping of the An from aluminium matrix into molten chloride media. The DOS process has been successfully demonstrated for treatment of oxide type fuels within the last years: the core of the process has been already assessed and the studies have shown high selectivity and a quantitative recovery of actinides. Within the framework of the SACSESS European research program, the pyrochemical activities focused on applications of the DOS process to reprocess CERCER transmutation targets. These particular types of fuels consist of a mixture of minor actinides (MA) oxides diluted in an inert (oxide MgO) matrices. The behaviour of matrices material was first investigated regarding the solubility in the fluoride salt, starting from both oxide powders or sintered pellets. The saturation of Mg in the salt could be estimated at ∼ 3 wt%. Regarding the reductive extraction, as expected no Mg was reduced by the metallic phase. The present work also highlights that Mg has low impact on the extraction efficiency of U as long as the salt is not saturated. Once the saturation occurs, the efficiency starts to decrease. So we recommend recycling the salt when Mg saturation is reached.

Proposed methods for treating high-level pyrochemical process wastes

International Nuclear Information System (INIS)

Johnson, T.R.; Miller, W.E.; Steunenberg, R.K.

1985-01-01

This survey illustrates the large variety and number of possible techniques available for treating pyrochemical wastes; there are undoubtedly other process types and many variations. The choice of a suitable process is complicated by the uncertainty as to what will be an acceptable waste form in the future for both TRU and non-TRU wastes

Pyrochemical recovery of actinide elements from spent light water reactor fuel

International Nuclear Information System (INIS)

Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

1994-01-01

Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl 2 salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl 2 transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product

Transportation of pyrochemical salts from Rocky Flats to Los Alamos

International Nuclear Information System (INIS)

Schreiber, S.B.

1997-01-01

Radioactive legacy wastes or residues are currently being stored on numerous Sites around the former Department of Energy's (DOE) Nuclear Weapons Complex. Since most of the operating facilities were shut down and have not operated since before the declared end to the Cold War in 1993, the historical method for treating these residues no longer exists. The risk associated with continued storage of these residues will dramatically increase with time. Thus, the DOE was directed by the Defense Nuclear Facility Safety Board in its Recommendation 94-1 to address and stabilize these residues and established an eight year time frame for doing so. There are only two options available to respond to this requirement: (1) restart existing facilities to treat and package the residues for disposal or (2) transport the residues to another operating facility within the Complex where they can be treated and packaged for disposal. This paper focuses on one such residue type, pyrochemical salts, produced at one Complex site, the Rocky Flats Plant located northwest of Denver, Colorado. One option for treating the salts is their shipment to Los Alamos, New Mexico, for handling at the Plutonium Facility. The safe transportation of these salts can be accomplished at present with several shipping containers including a DOT 6M, a DOE 9968, Type A or Type B quantity 55-gallon drum overpacks, or even the TRUPACT II. The tradeoffs between each container is examined with the conclusion that none of the available shipping containers is fully satisfactory. Thus, the advantageous aspects of each container must be utilized in an integrated and efficient way to effectively manage the risk involved. 1 fig

Pyrochemical head-end treatment for fast reactor fuel elements

International Nuclear Information System (INIS)

Avogadro, A.

1978-01-01

The paper presents the R and D work performed at Ispra and Mol during the period 1965-1975 in order to find a way to overcome technical and economical difficulties arising when the conventional reprocessing is applied to fast reactor fuel elements. The work had been directed towards 3 specific topics: a) liquid-metal decladding of spent stainless steel - clad fuels (solinox process). b) oxidative pulverisation by fused salts and extraction of volatile fission products (satex process). c) Pyrochemical separation of plutonium from the bulk of the fuel

Development of a vacuum distillation process for Pu pyro-chemistry spent salts treatment

International Nuclear Information System (INIS)

Bourges, G.; Lambertin, D.; Baudrot, C.; Pescayre, L.; Thiebaut, C.

2004-01-01

The pyrochemical purification of plutonium has generated spent salts, which are disposed in nuclear facility. To reduce stored quantities, the development of a pyrochemical treatment is in progress. The feed salt, typically composed of various Pu and Am species spread into monovalent or divalent chloride matrix, is first oxidized to convert the actinides into oxides. Then the chlorides are separated from the actinide oxides by vacuum distillation. Temperatures higher than 750 deg C for mono-chloride salts mixture NaCl/KCl and higher than 1100 deg C for divalent CaCl 2 base salts, are required to produce an industrial flow of vaporization. Inactive qualification of the process for NaCl/KCl base salt has been carried with lanthanide surrogates. Then, a pilot equipment, called Distillator has been designed and built for production-scale distillation of NaCl/KCl and CaCl 2 oxidized plutonium salts. Industrial flows of vaporization have been obtained with this pilot equipment: about 4 g/cm 2 /h for NaCl/KCl at 800 - 900 deg C and 1 Pa, and more than 1.5 g/cm 2 /h for CaCl 2 base salts between 1000 - 1200 deg C at 0.1 Pa. The last step will be the integration of the Distillator into a glove box. (authors)

Actinide recovery from pyrochemical residues

International Nuclear Information System (INIS)

Avens, L.R.; Clifton, D.G.; Vigil, A.R.

1984-01-01

A new process for recovery of plutonium and americium from pyrochemical waste has been demonstrated. It is based on chloride solution anion exchange at low acidity, which eliminates corrosive HCl fumes. Developmental experiments of the process flowsheet concentrated on molten salt extraction (MSE) residues and gave >95% plutonium and >90% americium recovery. The recovered plutonium contained 6 = from high chloride-low acid solution. Americium and other metals are washed from the ion exchange column with 1N HNO 3 -4.8M NaCl. The plutonium is recovered, after elution, via hydroxide precipitation, while the americium is recovered via NaHCO 3 precipitation. All filtrates from the process are discardable as low-level contaminated waste. Production-scale experiments are now in progress for MSE residues. Flow sheets for actinide recovery from electrorefining and direct oxide reduction residues are presented and discussed

Actinide recovery from pyrochemical residues

International Nuclear Information System (INIS)

Avens, L.R.; Clifton, D.G.; Vigil, A.R.

1985-05-01

We demonstrated a new process for recovering plutonium and americium from pyrochemical waste. The method is based on chloride solution anion exchange at low acidity, or acidity that eliminates corrosive HCl fumes. Developmental experiments of the process flow chart concentrated on molten salt extraction (MSE) residues and gave >95% plutonium and >90% americium recovery. The recovered plutonium contained 6 2- from high-chloride low-acid solution. Americium and other metals are washed from the ion exchange column with lN HNO 3 -4.8M NaCl. After elution, plutonium is recovered by hydroxide precipitation, and americium is recovered by NaHCO 3 precipitation. All filtrates from the process can be discardable as low-level contaminated waste. Production-scale experiments are in progress for MSE residues. Flow charts for actinide recovery from electro-refining and direct oxide reduction residues are presented and discussed

Development of quantitative analytical methods for the control of actinides in a pyrochemical partitioning process

International Nuclear Information System (INIS)

Abousahl, S.; Belle, P. van; Eberle, H.; Ottmar, H.; Lynch, B.; Vallet, P.; Mayer, K.; Ougier, M.

2005-01-01

Advanced nuclear fuel cycles are being developed in order to reduce the long-term radiotoxicity of highly radioactive waste. Pyrochemical partitioning techniques appear particularly attractive for advanced fuel cycles in which the minor actinides are recycled. The electrochemical processes of practical importance are the electrorefining process and the liquid-liquid extraction of transuranic (TRU) elements from fission products using either non-miscible molten metal or molten salt-metal phases. Analytical methods for the accurate assay of actinide elements in these matrices needed to be developed. A quantitative assay is required in order to establish a material balance for process development and - at a later stage - for accountancy and control purposes. To this end radiometric techniques such as energy-dispersive X-ray fluorescence analysis (XRF), neutron coincidence counting (NCC) and high-resolution gamma spectrometry (HRGS) were extensively employed for the quantitative determination of actinides (U, Np, Pu, Am, Cm) in process samples. Comparative analyses were performed using inductively coupled plasma mass spectrometry (ICP-MS). The respective samples were available in small quantities (≅ 100 mg) either in the form of eutectic salt or in metallic form with Cd, Zr or Bi as major metallic matrix constituents. (orig.)

Continuous extraction of molten chloride salts with liquid cadmium alloys

International Nuclear Information System (INIS)

Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.

1993-01-01

A pyrochemical method is being developed at Argonne National Laboratory (ANL) to provide contnuous multistage extractions between molten chloride salts and liquid cadmium alloys at 500 degrees C. The extraction method will be used to recover transuranic (TRU) elements from the process salt in the electroretiner used in the pyrochemical reprocessing of spent fuel from the Integral Fast Reactor (IFR). The IFR is one of the Department of Energy's advanced power reactor concepts. The recovered TRU elements are returned to the electrorefiner. The extracted salt undergoes further processing to remove rare earths and other fission products so that most of the purified salt can also be returned to the electrorefiner, thereby extending the useful life of the process salt many times

Molten salt/metal extractions for recovery of transuranic elements

International Nuclear Information System (INIS)

Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.

1992-01-01

The integral fast reactor (EFR) is an advanced reactor concept that incorporates metallic driver and blanket fuels, an inherently safe, liquid-sodium-cooled, pool-type, reactor design, and on-site pyrochemical reprocessing (including electrorefining) of spent fuels and wastes. This paper describes a pyrochemical method that is being developed at Argonne National Laboratory to recover transuranic elements from the EFR electrorefiner process salt. The method uses multistage extractions between molten chloride salts and cadmium metal at high temperatures. The chemical basis of the salt extraction method, the test equipment, and a test plan are discussed

Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel

International Nuclear Information System (INIS)

Boussier, H.; Heuer, D.

2010-01-01

The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Reactor Fast (MSFR).

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

International Nuclear Information System (INIS)

Ikegami, Tetsuo

2008-01-01

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle. (author)

Pyrochemical processing automation at Lawrence Livermore National Laboratory

International Nuclear Information System (INIS)

Dennison, D.K.; Domning, E.E.; Seivers, R.

1991-01-01

Lawrence Livermore National Laboratory (LLNL) is developing a fully automated system for pyrochemical processing of special nuclear materials (SNM). The system utilizes a glove box, an automated tilt-pour furnace (TPF), an IBM developed gantry robot, and specialized automation tooling. All material handling within the glove box (i.e., furnace loading, furnace unloading, product and slag separation, and product packaging) is performed automatically. The objectives of the effort are to increase process productivity, decrease operator radiation, reduce process wastes, and demonstrate system reliability and availability. This paper provides an overview of the automated system hardware, outlines the overall operations sequence, and discusses the current status

Uranium transport to solid electrodes in pyrochemical reprocessing of nuclear fuel

International Nuclear Information System (INIS)

Tomczuk, Z.; Ackerman, J.P.; Wolson, R.D.; Miller, W.E.

1992-01-01

A unique pyrochemical process developed for the separation of metallic nuclear fuel from fission products by electrotransport through molten LiCl-KCl eutectic salt to solid and liquid metal cathodes. The process allow for recovery and reuse of essentially all of the actinides in spent fuel from the integral fast reactor (IFR) and disposal of wastes in satisfactory forms. Electrotransport is used to minimize reagent consumption and, consequently, waste volume. In particular, electrotransport to solid cathodes is used for recovery of an essentially pure uranium product in the presence of other actinides; removal of pure uranium is used to adjust the electrolyte composition in preparation for recovery of a plutonium-rich mixture with uranium in liquid cadmium cathodes. This paper presents experiments that delineate the behavior of key actinide and rare-earth elements during electrotransport to a solid electrode over a useful range of PuCl 3 /UCl 3 ratios in the electrolyte, a thermodynamic basis for that behavior, and a comparison of the observed behavior with that calculated from a thermodynamic model. This work clearly established that recovery of nearly pure uranium can be a key step in the overall pyrochemical-fuel-processing strategy for the IFR

Treatment of High-Level Waste Arising from Pyrochemical Processes

International Nuclear Information System (INIS)

Lizin, A.A.; Kormilitsyn, M.V.; Osipenko, A.G.; Tomilin, S.V.; Lavrinovich, Yu.G.

2013-01-01

JSC “SSC RIAR” has been performing research and development activities in support of closed fuel cycle of fast reactor since the middle of 1960s. Fuel cycle involves fabrication and reprocessing of spent nuclear fuel (SNF) using pyrochemical methods of reprocessing in molten alkali metal chlorides. At present pyrochemical methods of SNF reprocessing in molten chlorides has reached such a level in their development that makes it possible to compare their competitiveness with classic aqueous methods. Their comparative advantage lies in high safety, compactness, high protectability as to nonproliferation of nuclear materials, and reduction of high level waste volume

Estimation of zirconium in various process streams in molten salt electrorefining process

International Nuclear Information System (INIS)

Suganthi, S.; Vandarkuzhali, S.; Venkatesh, P.; Prabhakara Reddy, B.; Nagarajan, K.

2012-01-01

Molten salt electrorefining process is a non-aqueous pyrochemical process suitable for reprocessing spent metallic fuel. In this process the spent fuel is taken at the anode and the fuel elements are selectively electrotransported to a suitable cathode (either a solid steel cathode or liquid cadmium cathode) using molten LiCl-KCI as electrolyte. We have demonstrated electrorefining of UZr alloy at engineering scale level. 1 Kg U-6%Zr alloy was taken at the anode and pure uranium was recovered at a steel cathode using molten LiCIKCI-5%UCI 3 as electrolyte at 773 K. In this paper we present the method of dissolution, sample preparation and estimation of zirconium in various process streams in the electrorefining experiments carried out in our laboratory

Pyrochemical and Dry Processing Methods Program. A selected bibliography

Energy Technology Data Exchange (ETDEWEB)

McDuffie, H.F.; Smith, D.H.; Owen, P.T.

1979-03-01

This selected bibliography with abstracts was compiled to provide information support to the Pyrochemical and Dry Processing Methods (PDPM) Program sponsored by DOE and administered by the Argonne National Laboratory. Objectives of the PDPM Program are to evaluate nonaqueous methods of reprocessing spent fuel as a route to the development of proliferation-resistant and diversion-resistant methods for widespread use in the nuclear industry. Emphasis was placed on the literature indexed in the ERDA--DOE Energy Data Base (EDB). The bibliography includes indexes to authors, subject descriptors, EDB subject categories, and titles.

Pyrochemical and Dry Processing Methods Program. A selected bibliography

International Nuclear Information System (INIS)

McDuffie, H.F.; Smith, D.H.; Owen, P.T.

1979-03-01

This selected bibliography with abstracts was compiled to provide information support to the Pyrochemical and Dry Processing Methods (PDPM) Program sponsored by DOE and administered by the Argonne National Laboratory. Objectives of the PDPM Program are to evaluate nonaqueous methods of reprocessing spent fuel as a route to the development of proliferation-resistant and diversion-resistant methods for widespread use in the nuclear industry. Emphasis was placed on the literature indexed in the ERDA--DOE Energy Data Base (EDB). The bibliography includes indexes to authors, subject descriptors, EDB subject categories, and titles

RIAR experimental base development concept 1. Multi-purpose pyrochemical complex for experimental justification of innovative closed fuel cycle technologies

Energy Technology Data Exchange (ETDEWEB)

Bychkov, A.V.; Kormilitsyn, M.V. [Research Institute of Atomic Reactors, Dimitrovgrad-10, Ulyanovsk region, 433510 (Russian Federation)

2009-06-15

The principles of closed FC arrangement on the basis of non-aqueous methods allow the development of production addressing two tasks simultaneously: production of fresh fuel and reprocessing of irradiated fuel, that makes it possible to achieve the industrial level of implementation of closed FC of fast reactors of new generation in a series variant of standardized process modules on the basis of innovative pyrochemical high-effective compact technologies. For the purpose of experimental justification of innovative closed FC technologies at the RIAR site, the existing experimental base is being updated and a multi-purpose pyrochemical complex is developed: - Experimental complex of pyrochemical molten salt facilities to reprocess all types of spent fuel (MOX, nitride, metallic, IMF) of fast reactors of new generation (BN-800, MBIR, BREST). - Experimental complex of facilities to master a gas-fluoride technology of reprocessing intractable fuel, research reactors fuel and thermal SNF. - Transition of the existing facility of pyro-electrochemical production of MOX fuel into the mode of reprocessing of the BN-800 MOX SNF. - Renovation of the facilities for production of fuel elements from experimental, re-fabricated, innovative and high-active fuel - a complex of heavy and glove boxes - to produce experimental fuel elements and targets with MAs on the basis of oxides (vibro and pellets), mixed nitrides, metal alloys and inert matrices in heavy boxes. - Upgrading of the complex for mastering and demonstration of the processes for radioactive waste management and spent fuel pyrochemical reprocessing. The report covers main concept and design solutions, plans and schedule of the program for development of pyrochemical complex for experimental justification of innovative closed FC technologies. (authors)

Conversion of metal plutonium to plutonium dioxide by pyrochemical method

Energy Technology Data Exchange (ETDEWEB)

Panov, A.V.; Subbotin, V.G. [Russian Federal Nuclear Center, ALL-Russian Science and Research Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

2000-07-01

Report contains experimental results on metal plutonium of weapon origin samples conversion to plutonium dioxide by pyrochemical method. Circuits of processes are described. Their advantages and shortcomings are shown. Parameters of plutonium dioxide powders (phase and fraction compositions, poured density) manufactured by pyrochemical method in RFNC-VNIITF are shown as well. (authors)

Pyrochemical head-end treatment for spent nuclear fuels

International Nuclear Information System (INIS)

Bowersox, D.F.

1977-01-01

A program based upon thermodynamic values and scouting experiments at Argonne National Laboratory is proposed for development of a pyrochemical head-end treatment of spent nuclear fuels to replace the proposed chopping and leaching operation in the Purex process. The treatment consists of separation of the cladding from the oxide fuel by dissolution into liquid zinc; oxide reduction of uranium and plutonium and dissolution into a zinc--magnesium alloy; separation of alkali, alkaline earth, and rare earth fission products into a molten salt; and, finally, separation and recovery of the plutonium and uranium in the alloy. Uranium and plutonium would be separated from the fuel cladding and selected fission products in a form readily dissolvable in nitric acid. The head-end process could be developed eventually into an optimum method for recovering uranium, plutonium, and selected fission products and for minimizing wastes as compact, stable solids. Developmental expenses are not known clearly, but the potential advantages of the process are impressive

Pyrochemical properties of actinides elements

International Nuclear Information System (INIS)

Akabori, Mitsuo; Hayashi, Hirokazu; Minato, Kazuo

2005-01-01

The present status of research works on pyrochemical processing is introduced in view of: 1) the process concept; 2) R and D progress; 3) new experimental facilities in NUCEF. Basic laboratory studies with rare-earth elements have almost been completed and hot tests with americium have been started at the TRU-HITEC facility in NUCEF. (author)

On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

International Nuclear Information System (INIS)

Vasin, B.D.; Ivanov, V.A.; Shchetinskij, A.V.; Vavilov, S.K.; Savochkin, Yu.P.; Bychkov, A.V.; Kormilitsyn, M.V.

2005-01-01

A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements [ru

Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

International Nuclear Information System (INIS)

Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

2007-01-01

Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

High-temperature vacuum distillation separation of plutonium waste salts

International Nuclear Information System (INIS)

Garcia, E.

1996-01-01

In this task, high-temperature vacuum distillation separation is being developed for residue sodium chloride-potassium chloride salts resulting from past pyrochemical processing of plutonium. This process has the potential of providing clean separation of the salt and the actinides with minimal amounts of secondary waste generation. The process could produce chloride salt that could be discarded as low-level waste (LLW) or low actinide content transuranic (TRU) waste, and a concentrated actinide oxide powder that would meet long-term storage standards (DOE-DTD-3013-94) until a final disposition option for all surplus plutonium is chosen

Pyrochemical reprocessing of nitride fuel

International Nuclear Information System (INIS)

Nakazono, Yoshihisa; Iwai, Takashi; Arai, Yasuo

2004-01-01

Electrochemical behavior of actinide nitrides in LiCl-KCl eutectic melt was investigated in order to apply pyrochemical process to nitride fuel cycle. The electrode reaction of UN and (U, Nd)N was examined by cyclic voltammetry. The observed rest potential of (U, Nd)N depended on the equilibrium of U 3+ /UN and was not affected by the addition of NdN of 8 wt.%. (author)

Interfacing solvent extraction in the recovery of pyrochemical residues at the Savannah River Plant

International Nuclear Information System (INIS)

Gray, L.W.; Holcomb, H.P.

1986-01-01

The traditional feedstock for plutonium recovery at the Savannah River Plant (SRP) has been spent reactor fuel elements and irradiated targets. Feed sources have included both onsite reactors and a wide variety of domestic and foreign reactors. For the past few years, a growing and increasingly varied mix of unirradiated plutonium residues has been purified through SRP aqueous-based processes. Recently, plutonium residues generated in various chloride salt melts have become a significant offsite source of feed for SRP recovery operations. Impure plutonium metal and plutonium alloys have also been processed. A broader range of molten salt and other high temperature residues is anticipated for the future. The major advantage of solvent extraction for scrap purification is the versatility of the solvent extraction system which allows numerous contaminants to be removed by routine operations. Major concerns are nuclear safety control, corrosion of equipment, and control of releases to the environment. SRP's past, present, and future interfacing of solvent extraction in processing pyrochemical and other plutonium-containing residues is reviewed

Pyrochemical treatment of Idaho Chemical Processing Plant high-level waste calcine

International Nuclear Information System (INIS)

Todd, T.A.; DelDebbio, J.A.; Nelson, L.O.; Sharpsten, M.R.

1993-01-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-level waste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described

Fluorescence Resonance Energy Transfer of the Tb(III)-Nd(III) Binary System in Molten LiCl-KCl Eutectic Salt

Energy Technology Data Exchange (ETDEWEB)

Kim, B. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yun, J. I. [KAIST, Daejeon (Korea, Republic of)

2015-05-15

The lanthanides act as a neutron poison in nuclear reactor with large neutron absorption cross section. For that reason, very low amount of lanthanides is required in the recovered U/TRU ingot product from pyrochemical process. In view of that, the investigation of thermodynamic properties and chemical behaviors of lanthanides in molten chloride salt are necessary to estimate the performance efficiency of pyrochemical process. However, there are uncertainties about knowledge and understanding of basic mechanisms in pyrochemical process, such as chemical speciation and redox behaviors due to the lack of in-situ monitoring methods for high temperature molten salt. The spectroscopic analysis is one of the probable techniques for in-situ qualitative and quantitative analysis. Recently, a few fluorescence spectroscopic measurements on single lanthanide element in molten LiCl-KCl eutectic have been investigated. The fluorescence intensity and the fluorescence lifetime of Tb(III) were decreased as increasing the concentration of Nd(III), demonstrating collisional quenching between donor ions and acceptor ions. The Forster distance (..0) of Tb(III)-Nd(III) binary system in molten LiCl-KCl eutectic was determined in the specific range of .... (0.1-1.0) and .. (1.387-1.496)

Waste Minimization Study on Pyrochemical Reprocessing Processes

International Nuclear Information System (INIS)

Boussier, H.; Conocar, O.; Lacquement, J.

2006-01-01

Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal

Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Science.gov (United States)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Development of High Temperature Transport System for Molten Salt

International Nuclear Information System (INIS)

Lee, S. H.; Lee, H. S.; Kim, J. G.

2011-01-01

Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes which is composed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyrometallurgical processing, the development of high-temperature molten salt transport technologies is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature transport technology for molten salt, and the performance test of the apparatus was performed. And also, predissolution test of the salt was carried out using the reactor with furnace in experimental apparatus

Activity coefficients of plutonium and cerium in liquid gallium at 1073 K: Application to a molten salt/solvent metal separation concept

International Nuclear Information System (INIS)

Lambertin, David; Ched'homme, Severine; Bourges, Gilles; Sanchez, Sylvie; Picard, Gerard S.

2005-01-01

Activity coefficients in liquid metal and salt phases are important parameters for predicting the separation efficiency of reductive extraction or electrochemical pyrochemical processes. The electrochemical properties of Ce and Pu in gallium metal and chlorides media - CaCl 2 and equimolar NaCl-KCl - have been studied at 1073 K. Cyclic voltammetry and chronoamperometry show the thermodynamic feasibility of using gallium as solvent metal for pyrochemical processes involving Pu and Ce. The activity coefficient of Pu in liquid Ga (log(γ Pu,Ga ) = -7.3 ± 0.5) is deduced from the results and is a basis of assessing the potential for using liquid metals in pyrochemical separation of actinides and lanthanides. Evaluation of literature data for Al, Bi and Cd suggests that Ga is most favorable for selective separation of Pu from Ce near 1073 K

Preliminary Study on the High Temperature Transport System for Molten Salt

International Nuclear Information System (INIS)

Lee, S. H.; Lee, H. S.; Kim, J. G.

2012-01-01

Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes is compos- ed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyroprocessing technology, the development of high-temperature transport technologies for molten salt is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt

Evaluation of in-plant neutron coincidence counters for the measurement of molten salt extraction residues

International Nuclear Information System (INIS)

Langner, D.G.; Russo, P.A.; Wachter, J.R.

1993-01-01

Americium is extracted from plutonium by a molten salt extraction (MSE) process. The residual americium-laden salts are a significant waste stream in this pyrochemical purification process. Rapid assay of MSE residues is desirable to minimize the exposure of personnel to these often high-level emissions. However, the quantitative assay of plutonium in MSE residues is difficult. Variable, unknown (a,n) rates and variable emitted-neutron energy spectra preclude the use of standard neutron coincidence counting techniques with old-generation neutron coincidence counters. Gamma-ray assay methods have not been successful with some residues because of random lumps of plutonium metal. In this paper, we present measurements of MSE residues with two state-of-the-art neutron coincidence counters at the Los Alamos Plutonium Processing Facility: an in-line counter built for the assay of bulk waste material and the pyrochemical multiplicity counter that underwent test and evaluation at that facility. Both of these counters were designed to minimize the effects on measurements of variations in the sample geometry and variable energy spectra of emitted neutrons. These results are compared to measurements made with an HLNCII and with a 20-yr-old in-line well counter. The latter two counters are not optimized in ft sense. We conclude that the newer counters provide significantly improved assay results. The pyrochemical multiplicity counter operated in the conventional coincidence mode provided the best assays overall

Zone Freezing Study for Pyrochemical Process Waste Minimization

Energy Technology Data Exchange (ETDEWEB)

Ammon Williams

2012-05-01

Pyroprocessing technology is a non-aqueous separation process for treatment of used nuclear fuel. At the heart of pyroprocessing lies the electrorefiner, which electrochemically dissolves uranium from the used fuel at the anode and deposits it onto a cathode. During this operation, sodium, transuranics, and fission product chlorides accumulate in the electrolyte salt (LiCl-KCl). These contaminates change the characteristics of the salt overtime and as a result, large volumes of contaminated salt are being removed, reprocessed and stored as radioactive waste. To reduce the storage volumes and improve recycling process for cost minimization, a salt purification method called zone freezing has been proposed at Korea Atomic Energy Research Institute (KAERI). Zone freezing is melt crystallization process similar to the vertical Bridgeman method. In this process, the eutectic salt is slowly cooled axially from top to bottom. As solidification occurs, the fission products are rejected from the solid interface and forced into the liquid phase. The resulting product is a grown crystal with the bulk of the fission products near the bottom of the salt ingot, where they can be easily be sectioned and removed. Despite successful feasibility report from KAERI on this process, there were many unexplored parameters to help understanding and improving its operational routines. Thus, this becomes the main motivation of this proposed study. The majority of this work has been focused on the CsCl-LiCl-KCl ternary salt. CeCl3-LiCl-KCl was also investigated to check whether or not this process is feasible for the trivalent species—surrogate for rare-earths and transuranics. For the main part of the work, several parameters were varied, they are: (1) the retort advancement rate—1.8, 3.2, and 5.0 mm/hr, (2) the crucible lid configurations—lid versus no-lid, (3) the amount or size of mixture—50 and 400 g, (4) the composition of CsCl in the salt—1, 3, and 5 wt%, and (5) the

Development of the object-oriented analysis code for the estimation of material balance in pyrochemical reprocessing process (2). Modification of the code for the analysis of holdup of nuclear materials in the process

International Nuclear Information System (INIS)

Okamura, Nobuo; Tanaka, Hiroshi

2001-04-01

Pyrochemical reprocessing is thought to be promising process for FBR fuel cycle mainly from the economical viewpoint. However, the material behavior in the process is not clarified enough because of the lack of experimental data. The authors have been developed the object-oriented analysis code for the estimation of material balance in the process, which has the flexible applicability for the change of process flow sheet. The objective of this study is to modify the code so as to analyze the holdup of nuclear materials in the pyrochemical process from the viewpoint of safeguard, because of the possibility of larger amount of the holdup in the process compared with aqueous process. As a result of the modification, the relationship between the production of nuclear materials and its holdup in the process can be evaluated by the code. (author)

Use of porous MgO in pyrochemical applications

International Nuclear Information System (INIS)

Maiya, P.S.; Sweeney, S.M.; Carroll, L.A.; Dusek, J.T.

1994-11-01

Pyrochemical methods for the extraction of transuranic elements from light water reactor spent fuel require a reduction step in which the oxide fuel is reduced to metals by Li in molten LiCl. The Li 2 O formed is electrolytically reduced to metal in a cell that uses a carbon (or inert) anode and a Li cathode to recycle the salt and minimize the waste. Use of a carbon anode causes carbon dust that interferes with the process. Moreover, current efficiency is reduced as a result of oxidation of Li to Li 2 O by CO 2 . A porous MgO shroud around the anode was found to obviate these problems. Porous MgO crucibles and rectangular bar specimens were fabricated from MgO powders (electrically fused MgO, reagent grade MgO were mixed in appropriate combinations with a binder and lubricant). Particle size, force applied to the powders during cold pressing, and sintering temperature were varied to achieve a total porosity of >45% (mostly open porosity) and to control pore size and pore distribution. Mercury intrusion porosimetry was used to determine the pore size and pore size distribution. Flexural strength is observed to be proportional to the square root of pore size, which is consistent with fracture mechanics

A reactive distillation process for the treatment of LiCl-KCl eutectic waste salt containing rare earth chlorides

Energy Technology Data Exchange (ETDEWEB)

Eun, H.C., E-mail: ehc2004@kaeri.re.kr; Choi, J.H.; Kim, N.Y.; Lee, T.K.; Han, S.Y.; Lee, K.R.; Park, H.S.; Ahn, D.H.

2016-11-15

The pyrochemical process, which recovers useful resources (U/TRU metals) from used nuclear fuel using an electrochemical method, generates LiCl-KCl eutectic waste salt containing radioactive rare earth chlorides (RECl{sub 3}). It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste salt in a hot-cell facility. For this reason, a reactive distillation process using a chemical agent was achieved as a method to separate rare earths from the LiCl-KCl waste salt. Before conducting the reactive distillation, thermodynamic equilibrium behaviors of the reactions between rare earth (Nd, La, Ce, Pr) chlorides and the chemical agent (K{sub 2}CO{sub 3}) were predicted using software. The addition of the chemical agent was determined to separate the rare earth chlorides into an oxide form using these equilibrium results. In the reactive distillation test, the rare earth chlorides in LiCl-KCl eutectic salt were decontaminated at a decontamination factor (DF) of more than 5000, and were mainly converted into oxide (Nd{sub 2}O{sub 3}, CeO{sub 2}, La{sub 2}O{sub 3}, Pr{sub 2}O{sub 3}) or oxychloride (LaOCl, PrOCl) forms. The LiCl-KCl was purified into a form with a very low concentration (<1 ppm) for the rare earth chlorides.

Recovery of Residual LiCl-KCl Eutectic Salts in Radioactive Rare Earth Precipitates

International Nuclear Information System (INIS)

Eun, Hee Chul; Yang, Hee Chul; Kim, In Tae; Lee, Han Soo; Cho, Yung Zun

2010-01-01

For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily

Numerical Modelling of Induction Heating for a Molten Salts Pyrochemical Process

Energy Technology Data Exchange (ETDEWEB)

Vu, Xuan-Tuyen; Feraud, Jean-Pierre; Ode, Denis [CEA Marcoule: DTEC/SGCS/LGCI Bat. 57 B17171, 30207 Bagnols/Ceze (France); Du Terrail Couvat, Yves [SIMaP, Grenoble INP, CNRS: ENSEEG, BP 75, 38402 Saint Martin d' Heres Cedex (France)

2008-07-01

Technological developments in the pyro-chemistry program are required to allow choices for a reprocessing experiment on 100 g of spent nuclear fuel. In this context, a special device must be designed for the solid/gas reaction phases followed by actinide extraction and stripping in molten salt. This paper discusses a modelling approach for designing an induction furnace. Using this numerical approach is a good way to improve thermal performance of the device in terms of magnetic/thermal coupling phenomena. The influence of current frequency is also studied to give another view of the possibilities of an induction furnace. Electromagnetic forces are taken into account in a computational fluid dynamics code derived from a specifically developed exchange library. Induction heating systems are an example of a typical multi-physics problem involving numerically coupled equations. (authors)

Numerical Modelling of Induction Heating for a Molten Salts Pyrochemical Process

International Nuclear Information System (INIS)

Vu, Xuan-Tuyen; Feraud, Jean-Pierre; Ode, Denis; Du Terrail Couvat, Yves

2008-01-01

Technological developments in the pyro-chemistry program are required to allow choices for a reprocessing experiment on 100 g of spent nuclear fuel. In this context, a special device must be designed for the solid/gas reaction phases followed by actinide extraction and stripping in molten salt. This paper discusses a modelling approach for designing an induction furnace. Using this numerical approach is a good way to improve thermal performance of the device in terms of magnetic/thermal coupling phenomena. The influence of current frequency is also studied to give another view of the possibilities of an induction furnace. Electromagnetic forces are taken into account in a computational fluid dynamics code derived from a specifically developed exchange library. Induction heating systems are an example of a typical multi-physics problem involving numerically coupled equations. (authors)

Development of fuel cycle technology for molten-salt reactor systems

International Nuclear Information System (INIS)

Uhlir, J.

2006-01-01

Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

Spectro-electrochemical Studies of Europium and Uranium Ions in LiCl-KCl Eutectic

Energy Technology Data Exchange (ETDEWEB)

Bae, Sang Eun; Park, Yong Joon; Cho, Young Hwan; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-05-15

Pyrochemical processing of nuclear fuels using a molten salt as a solvent is regarded as one of the promising options for future spent nuclear fuel management. Molten salts are known as suitable media for electrorefining and electrowinning of metal in the Pyrochemical process. There are complicated chemical and electrochemical reactions in the molten salt of the Pyrochemical process. In order to reach a better understanding and control of these metal deposition processes, accurate knowledge of the reaction mechanism is essential. Spectroscopic methods, such as fluorescence and UV-VIS spectroscopy, are considered to be one of powerful tools to investigate the chemical elements and its oxidation state. In this work, the spectroscopic studies have been performed under the electrochemical control to investigate the reaction mechanisms in the molten salt at high temperature during the electrochemical reactions

Spectro-electrochemical Studies of Europium and Uranium Ions in LiCl-KCl Eutectic

International Nuclear Information System (INIS)

Bae, Sang Eun; Park, Yong Joon; Cho, Young Hwan; Song, Kyu Seok

2010-01-01

Pyrochemical processing of nuclear fuels using a molten salt as a solvent is regarded as one of the promising options for future spent nuclear fuel management. Molten salts are known as suitable media for electrorefining and electrowinning of metal in the Pyrochemical process. There are complicated chemical and electrochemical reactions in the molten salt of the Pyrochemical process. In order to reach a better understanding and control of these metal deposition processes, accurate knowledge of the reaction mechanism is essential. Spectroscopic methods, such as fluorescence and UV-VIS spectroscopy, are considered to be one of powerful tools to investigate the chemical elements and its oxidation state. In this work, the spectroscopic studies have been performed under the electrochemical control to investigate the reaction mechanisms in the molten salt at high temperature during the electrochemical reactions

Nonaqueous processing methods

International Nuclear Information System (INIS)

Coops, M.S.; Bowersox, D.F.

1984-09-01

A high-temperature process utilizing molten salt extraction from molten metal alloys has been developed for purification of spent power reactor fuels. Experiments with laboratory-scale processing operations show that purification and throughput parameters comparable to the Barnwell Purex process can be achieved by pyrochemical processing in equipment one-tenth the size, with all wastes being discharged as stable metal alloys at greatly reduced volume and disposal cost. This basic technology can be developed for large-scale processing of spent reactor fuels. 13 references, 4 figures

Technical-and-economic analysis and optimization of the full flow charts of processing of radioactive wastes on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of fast reactors

Science.gov (United States)

Gupalo, V. S.; Chistyakov, V. N.; Kormilitsyn, M. V.; Kormilitsyna, L. A.; Osipenko, A. G.

2015-12-01

When considering the full flow charts of processing of radioactive wastes (RAW) on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of NIIAR fast reactors, we corroborate optimum technical solutions for the preparation of RAW for burial from a standpoint of heat release, dose formation, and technological storage time with allowance for technical-and-economic and ecological indices during the implementation of the analyzed technologies and equipment for processing of all RAW fluxes.

Development of interface technology between unit processes in E-Refining process

Energy Technology Data Exchange (ETDEWEB)

Lee, S. H.; Lee, H. S.; Kim, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

The pyroprocessing is composed mainly four subprocesses, such as an electrolytic reduction, an electrorefining, an electrowinning, and waste salt regeneration/ solidification processes. The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, is under development by Korea Atomic Energy Research Institute as a sub process of pyrochemical treatment of spent PWR fuel. The CERS(Continuous ElectroRefining System) is composed of some unit processes such as an electrorefiner, a salt distiller, a melting furnace for the U-ingot and U-chlorinator (UCl{sub 3} making equipment) as shown in Fig. 1. In this study, the interfaces technology between unit processes in E-Refining system is investigated and developed for the establishment of integrated E-Refining operation system as a part of integrated pyroprocessing

Densification of salt-occluded zeolite a powders to a leach-resistant monolith

International Nuclear Information System (INIS)

Lewis, M.A.; Fischer, D.F.; Murhpy, C.D.

1993-01-01

Pyrochemical processing of spent fuel from the Integral Fast Reactor (IFR) yields a salt waste of LiCl-KCl that contains approximately 6 wt% fission products, primarily as CsCl and SrCl 2 . Past work has shown that zeolite A will preferentially sorb cesium and strontium and will encapsulate the salt waste in a leach-resistant, radiation-resistant aluminosilicate matrix. However, a method is sill needed to convert the salt-occluded zeolite powders into a form suitable for geologic disposal. We are thus investigating a method that forms bonded zeolite by hot pressing a mixture of glass frit and salt-occluded zeolite powders at 990 K (717 degree C) and 28 MPa. The leach resistance of the bonded zeolite was measured in static leach tests run for 28 days in 363 K (90 degree C) deionized water. Normalized release rates of all elements in the bonded zeolite were low, 2 d. Thus, the bonded zeolite may be a suitable waste form for IFR salt waste

IFR electrorefining process

International Nuclear Information System (INIS)

Kim, S. B.

1997-01-01

A metallic fuel alloy, which is a key element of the IFR fuel cycle, permits the use of pyrochemical processing of the spent fuel. Electrorefining with molten salt electrolytes is a key step in the pyroprocess because the actinides are recovered, separated from the fission products present in the spent fuel in this operation and then recycled for use as fuel. Chemical and electrochemical aspects of the electrorefining method is to be described. (author)

Pyro-oxidation of plutonium spent salts with sodium carbonate

International Nuclear Information System (INIS)

Bourges, G.; Godot, A.; Valot, C.; Devillard, D.

2001-01-01

The purification of plutonium generates spent salts, which are temporarily stored in a nuclear building. A development programme for pyrochemical treatment is in progress to stabilize and concentrate these salts in order to reduce the quantities for long-term disposal. The treatment, inspired by work previously done by LANL, consists of a pyro-oxidation of the salt with sodium carbonate to convert the actinides into oxides, then of a vacuum distillation to separate the oxides from the volatile salt matrix. Pyro-oxidation of NaCl/KCl base spent salts first produces a 'black salt' which contains more than 97% of the initial actinides. XRD analyses indicate PuO 2 as major plutonium species and sodium plutonates or plutonium sub-oxides PuO 2-x can also be identified. Next appears a 'white salt' containing less than 500 ppm of plutonium, which meets the operational criterion for LLW discard. For these salts, the pyro-oxidation process in and of itself is expected to reduce the quantities to be stored on-site by more than one-third. The pyro-oxidation of CaCl 2 /NaCl base americium extraction salts leads to oxides PuO 2 and probably AmO 2 , but the yield of concentration in the black salt is lower and the white salt cannot be discarded as LLW. During vacuum distillation, excess carbonate can dissociate and damage the efficiency of the process. Appropriate chlorine sparging at the end of the oxidation can eliminate this carbonate. (authors)

R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

International Nuclear Information System (INIS)

Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

2006-01-01

The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

1994-01-01

Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

Detection of Fluorescence for Lanthanides in LiCl-KCl Molten Salt Medium

International Nuclear Information System (INIS)

Im, Hee Jung; Kim, Tack Jin; Song, Kyu Seok; Jee, Kwang Yong

2007-01-01

In the electrorefining step of the pyrochemical process, actinide ions dissolved in the LiCl-KCl eutectic salt are recovered as pure actinide metals at a cathode for a re-use as a nuclear fuel from the aspect of its nonproliferation of the nuclear fuel cycles. The lanthanide species dissolved in the LiCl-KCl eutectic salt play an important role in an effective metal purification during the electrorefining step, so it is necessary to understand the chemical and physical behaviors of lanthanides in molten salt. The in situ spectroscopic measurement system and studies according to temperature changes are essential for better understandable information. To our knowledge, the absorption studies of lanthanides at high temperatures have been reported before, but the fluorescence studies of those at high temperature are not reported yet. We will discuss here the fluorescence behaviors of lanthanides in LiCl-KCl molten salt medium according to a changing temperature

Interrelation of technologies for RW preparation and sites for final isolation of the wastes from pyrochemical processing of SNF

Energy Technology Data Exchange (ETDEWEB)

Gupalo, V.S.; Chistyakov, V.N. [JSC - Design-Prospecting and Scientific-Research Institute of Industrial Technology -, Kashirskoye Highway, 33, Moscow 115409 (Russian Federation); Kormilitsyn, M.V.; Kormilitsyna, L.A. [JSC - State Scientific Center - Research Institute of Atomic Reactors -, Ulyanovsk region, Dimitrovgrad - 10, 433510 (Russian Federation)

2013-07-01

For the justification of engineering solutions and practical testing of the radiochemical component of the perspective nuclear power complex with on-site variant of nuclear fuel cycle (NFC), it is planned to establish a multi-functional research-development complex (MFCRC) for radiochemical processing of spent nuclear fuels (SNF) from fast reactors. MFCRC is being established at the NIIAR site, it comprises technological process lines, where innovation pyro-electrochemical and hydrometallurgical technologies are realized, with an option for closing the inter-chain material flows for testing the combined radiochemically converted materials. The technological flowchart for processing at the MFCRC is subdivided into 3 segments: -) complex of the lead operations for dismantling the fuel elements (FE) and fuel assemblies (FA), -) pyrochemical extraction flowchart for processing SNF, and -) hydrometallurgical flowchart for processing SNF. The engineered solutions for the management and disposition of the radioactive wastes from MFCRC are reviewed.

Pyrophoric potential of plutonium-containing salt residues

International Nuclear Information System (INIS)

Haschke, John M.; Fauske, Hans K.; Phillips, Alan G.

2000-01-01

Ignition temperatures of plutonium and the pyrophoric potential of plutonium-containing pyrochemical salt residues are determined from differential thermal analysis (DTA) data and by modeling of thermal behavior. Exotherms observed at 90-200 deg. C for about 30% of the residues are attributed to reaction of plutonium with water from decomposition of hydrated salts. Exotherms observed near 300 deg. C are consistent with ignition of metal particles embedded in the salt. Onset of self-sustained reaction at temperatures as low as 90 deg. C is not precluded by these results and heat-balance models are developed and applied in predicting the static ignition point of massive metal and in evaluating salt pyrophoricity. Results show that ambient temperatures in excess of 200 deg. C are required for ignition of salt residues and that the most reactive salts cannot ignite at low temperatures because diffusion of oxidant to embedded metal is limited by low salt porosity

Status of the development on simulation technology for pyrochemical reprocessing

International Nuclear Information System (INIS)

Arie, Kazuo; Kawabe, Akihiro; Fujita, Reiko; Yamamura, Tsutomu; Sato, Yuzuru; Goto, Takuya; Minato, Kazuo; Tosaka, Ikuo; Amano, Osamu; Yamamoto, Kazuhiko

2005-01-01

The computerized simulation technique of element behaviors in pyrochemical reprocessing is largely useful to raise the efficiency in development of pyrochemical reprocessing technique and moreover to operate reasonably practical plans in the near future. The simulation code SPR1.0 has currently been developed, which can simultaneously analyze the electrochemical and chemical reactions and can deal generally also with many elements of study. It was found from some trial calculations that this code can analyze an electrolytic behavior of MOX. The present study was performed by Toshiba Co., Ltd. together with Tohoku University. and Kyoto University as entrusted from Japan Atomic Power Company Co., Ltd. in cooperation with JAERI, Tohoku Electric Power Co., Inc. and Tokyo Electric Power Co., Inc. (M.H.)

Stabilization of plutonium bearing residues at Lawrence Livermore National Laboratory

International Nuclear Information System (INIS)

Bronson, M.C.; Van Konynenburg, R.A.; Ebbinghaus, B.B.

1995-01-01

The US Department of Energy's (US DOE) Lawrence Livermore National Laboratory (LLNL) has plutonium holdings including metal, oxide and residue materials, all of which need stabilization of some type. Residue materials include calcined ash, calcined precipitates, pyrochemical salts, glove box sweepings, metallurgical samples, graphite, and pyrochemical ceramic crucibles. These residues are typical of residues stored throughout the US DOE plutonium sites. The stabilization process selected for each of these residues requires data on chemical impurities, physical attributes, and chemical forms of the plutonium. This paper outlines the characterization and stabilization of LLNL ash residues, pyrochemical salts, and graphite

Study of the actinide-lanthanide separation from nuclear waste by a new pyrochemical process

International Nuclear Information System (INIS)

Lemort, F.

1997-01-01

The theoretical extraction and separation of platinoids, actinides and lanthanides is allowed by thermodynamic using two adapted reducing agents: zinc and magnesium. Thereby, a pyrochemical method for the nuclear waste processing has been devised. The high temperature handling of the elements in fluoride forms and their processing by a reactive metallic phase required special precautions. The study of the behavior of matter in exploratory systems allowed the development of an experimental technology for the treatment and contacting of phases. The thermodynamical analysis of the experimental results shows the feasibility of the process. A model was developed to predict the distribution coefficients of zirconium, uranium and lanthanum as a function of the system composition. An estimation method was proposed in order to evaluate the distribution coefficients in diluted solution of all the actinides and lanthanides existing in the fission products between LiF CaF 2 and Zn-Mg at 720 deg C. Coupled with the experimental results, the estimates results may be extrapolated to concentrated solutions allowing predictions of the separation of all actinides and lanthanides. The rapidity of element transfer is induced by a thermal effect caused by the high exothermicity of the reduction by magnesium. The kinetic coefficients have been linked with the reduction enthalpy of each element. Moreover, the kinetics seem limited by chemical reaction and not by mass transfer. (author)

A study on recovery of uranium in the anode basket residues delivered from the pyrochemical process of used nuclear fuel

Science.gov (United States)

Eun, H. C.; Kim, T. J.; Jang, J. H.; Kim, G. Y.; Park, S. B.; Yoon, D. S.; Kim, S. H.; Paek, S. W.; Lee, S. J.

2018-04-01

In this study, the chlorination of uranium oxide (UO2) using ammonium chloride and zirconium as chemical agents was conducted to recover the uranium in the anode basket residues from the pyrochemical process of used nuclear fuel. The chlorination of UO2 was predicted using thermodynamic equilibrium calculations. The experimental conditions for the chlorination were determined using a chlorination test with cerium oxide (CeO2). In the chlorination test, it was confirmed that UO2 was chlorinated into UCl3 at 320 °C, some UO2 remained without changes in the chemical form, and ZrO2, Zr2O, and ZrCl2 were generated as byproducts.

Direct oxide reduction (DOR) solvent salt recycle in pyrochemical plutonium recovery operations

International Nuclear Information System (INIS)

Fife, K.W.; Bowersox, D.F.; Davis, C.C.; McCormick, E.D.

1987-02-01

One method used at Los Alamos for producing plutonium metal is to reduce the oxide with calcium metal in molten CaCl 2 at 850 0 C. The solvent CaCl 2 from this reduction step is currently discarded as low-level radioactive waste because it is saturated with the reaction by-product, CaO. We have developed and demonstrated a molten salt technique for rechlorinating the CaO, thereby regenerating the CaCl 2 and incorporating solvent recycle into the batch PuO 2 reduction process. We discuss results from the process development experiments and present our plans for incorporating the technique into an advanced design for semicontinuous plutonium metal production

Direct reduction of uranium dioxide and few other metal oxides to corresponding metals by high temperature molten salt electrolysis

International Nuclear Information System (INIS)

Mohandas, K.S.

2017-01-01

Molten salt based electro-reduction processes, capable of directly converting solid metal oxides to metals with minimum intermediate steps, are being studied worldwide. Production of metals apart, the process assumes importance in nuclear technology in the context of pyrochemical reprocessing of spent oxide fuels, for it serves as an intermediate step to convert spent oxide fuel to a metal alloy, which in turn can be processed by molten salt electro-refining method to gain the actinides present in it. In the context of future metal fuel fast reactor programme, the electrochemical process was studied for conversion of solid UO_2 to U metal in LiCl-1wt.% Li_2O melt at 650 °C with platinum anode at the Metal Processing Studies Section, PMPD, IGCAR. A brief overview of the work is presented in the paper

Status of development on simulation technology for pyrochemical reprocessing

International Nuclear Information System (INIS)

Arie, Kazuo; Mizuguchi, Koji; Fujita, Reiko

2004-01-01

Simulation techniques for the elemental behaviors in the pyrochemical reprocessing process of spent nuclear fuels are important for fuel reprocessing and future power station development. The authors developed a simulation code SPR1.0 which can analyze co-occurring electrochemical and chemical reactions simultaneously and which is applicable to know the behavior of any element in the system. The present report describes the status of the code development, the database for fundamental electrochemical reactions, and verification of the code. The code employs TRIAS code for electrochemical reactions and SOLGASMIX-PV for chemical reactions. Electrolytic process for MOX (mixed oxide) fuels with different Pu redox ratios were simulated using the present code and the effect of introducing iron ions was studied. The prospect of future development is also described. (S. Ohno)

Technical report for fabrication and performance test of electrochemical/spectroscopic measurement system

International Nuclear Information System (INIS)

Park, Yong Joon; Cho, Young Hwan; Bae, Sang Eun; Im, Hee Jung; Song, Kyu Seok

2010-01-01

Development of evaluation technology of electrochemical reactions is very essential to understand chemical behavior of actinides and lanthanides in molten salt media in relation to the development of Pyrochemical process. The on-line electrochemical/spectroscopic measurement system is to produce electrochemical parameters and thermodynamic parameters of actinides and lanthanides in molten salts by using spectroscopic techniques such as UV-VIS absorption as well as electrochemical in-situ measurement techniques. The on-line electrochemical/spectroscopic measurement system can be applied to understand the chemical reactions and oxidation states of actinides and lanthanides in molten salts eventually for the Pyrochemical process

PYRO, a system for modeling fuel reprocessing

International Nuclear Information System (INIS)

Ackerman, J.P.

1989-01-01

Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions

Proceedings of the workshop on molten salts technology and computer simulation

Energy Technology Data Exchange (ETDEWEB)

Hayashi, Hirokazu; Minato, Kazuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

Applications of molten salts technology to separation and synthesis of materials have been studied eagerly, which would develop new fields of materials science. Research Group for Actinides Science, Department of Materials Science, Japan Atomic Energy Research Institute (JAERI), together with Reprocessing and Recycle Technology Division, Atomic Energy Society of Japan, organized the Workshop on Molten Salts Technology and Computer Simulation at Tokai Research Establishment, JAERI on July 18, 2001. In the workshop eleven lectures were made and lively discussions were there on the fundamentals and applications of the molten salts technology that covered the structure and basic properties of molten salts, the pyrochemical reprocessing technology and the relevant computer simulation. The 10 of the presented papers are indexed individually. (J.P.N.)

Materials Science and Technology (MST) Division, Nuclear Materials Process Technology Group (MST-12), chemical process research and development report

International Nuclear Information System (INIS)

Clifton, D.G.

1984-04-01

A process for the recovery of plutonium and americium from molten salt extraction (MSE) salt residues has been demonstrated. It is based upon a new chloride anion-exchange process at low acidity that eliminates corrosive HCl fumes. The Los Alamos americium oxide production line has been improved to give more product with a concurrent lowering of personnel radiation exposure. A cost study has been made for the disposal of americium-contaminated calcium metal buttons that were obtained by pyrochemical recovery of plutonium from MSE salts. The waste form used in the study conforms to WIPP-Facility standards and current state-of-the-art radioactive waste disposal. The cost estimate is approx. $300/g 241 Am. Plutonium decontamination factors of approx. 300 have been obtained from lead-platinum alloy dissolution experiments carried out in alumina crucibles using lead oxide slag to getter the plutonium

Pyrochemical separation of radioactive components from inert materials in ICPP high-level calcined waste

International Nuclear Information System (INIS)

Del Debbio, J.A.; Nelson, L.O.; Todd, T.A.

1995-05-01

Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m 3 of high-level waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of 137 Cs and 90 Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details

New high-level waste management technology for IFR pyroprocessing wastes

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.

1993-01-01

The pyrochemical electrorefining process for recovery of actinides in spent fuel from the Integral Fast Reactor accumulates fission product wastes as chlorides dissolved in molten LiCI-KCI and as metals, some of which are in molten cadmium. Pyrochemical processes are being developed to recover uranium and transuranium elements for return to the reactor, and to separate and immobilize fission products in suitable waste forms. Solvent cadmium is recycled within the process. Electrolyte salt is treated in a series of salt/cadmium extraction steps; it is also returned to the process. Salt-borne fission products are concentrated on a zeolite bed that is converted to a stable, leach-resistant mineral. Rare earth fission products from the salt, noble metal fission products, and cladding hulls are dispersed in a metal matrix

Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Allen, Todd [Univ. of Wisconsin, Madison, WI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Simpson, Mike [Idaho National Lab., (United States)

2012-11-30

This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal.

Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

International Nuclear Information System (INIS)

Sridharan, Kumar; Allen, Todd; Anderson, Mark; Simpson, Mike

2012-01-01

This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal

Electrodeposition of alkali and alkali-earth metals on liquid lead cathodes in molten salts

International Nuclear Information System (INIS)

Caravaca, C.; De Cordoba, G.

2008-01-01

Pyrochemical processing of spent nuclear fuel leads to the dissolution as chlorides of fission products (FPs) that have to be removed in order to recycle the salt. Precipitation technique have been tested for the removal of these FPs in the LiCl-KCl, salt selected as reference, with different results. Salt decontamination from lanthanides can be easily achieved as solid precipitates of oxychlorides or single phosphates; however, for the alkaline and alkaline-earth metals this technique is not suitable. Within the EUROPART project (VI FP of the EC), a new route that consist of the electrodeposition of these FP on a liquid lead cathode (LLC) has been considered, including the Li and K constituting the electrolyte. First results obtained with Sr and Cs are presented herein. Although according to the thermodynamic potential values, the electrodeposition order on LLC is Ba, Sr, Li, K and Cs, during our experiments it was not possible to distinguish the electrochemical signals corresponding to the individual elements. (authors)

Electrodeposition of alkali and alkali-earth metals on liquid lead cathodes in molten salts

Energy Technology Data Exchange (ETDEWEB)

Caravaca, C.; De Cordoba, G. [CIEMAT/DE/DFN/URAA. Avda. Complutense, 22. 28040 Madrid (Spain)

2008-07-01

Pyrochemical processing of spent nuclear fuel leads to the dissolution as chlorides of fission products (FPs) that have to be removed in order to recycle the salt. Precipitation technique have been tested for the removal of these FPs in the LiCl-KCl, salt selected as reference, with different results. Salt decontamination from lanthanides can be easily achieved as solid precipitates of oxychlorides or single phosphates; however, for the alkaline and alkaline-earth metals this technique is not suitable. Within the EUROPART project (VI FP of the EC), a new route that consist of the electrodeposition of these FP on a liquid lead cathode (LLC) has been considered, including the Li and K constituting the electrolyte. First results obtained with Sr and Cs are presented herein. Although according to the thermodynamic potential values, the electrodeposition order on LLC is Ba, Sr, Li, K and Cs, during our experiments it was not possible to distinguish the electrochemical signals corresponding to the individual elements. (authors)

The process system analysis for advanced spent fuel management technology (I)

International Nuclear Information System (INIS)

Lee, H. H.; Lee, J. R.; Kang, D. S.; Seo, C. S.; Shin, Y. J.; Park, S. W.

1997-12-01

Various pyrochemical processes were evaluated, and viable options were selected in consideration of the proliferation safety, technological feasibility and compatibility to the domestic nuclear power system. Detailed technical analysis were followed on the selected options such as unit process flowsheet including physico-chemical characteristics of the process systems, preliminary concept development, process design criteria and materials for equipment. Supplementary analysis were also carried out on the support technologies including sampling and transport technologies of molten salt, design criteria and equipment for glove box systems, and remote operation technologies. (author). 40 refs., 49 tabs., 37 figs

Pyrochemical separations technologies envisioned for the U.S. accelerator transmutation of waste system

International Nuclear Information System (INIS)

Laidler, J. J.

2000-01-01

A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The baseline process selected combines aqueous and pyrochemical processes to enable the efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel. The diversity of processing methods was chosen for both technical and economic factors. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system

Laser-Induced Breakdown Spectroscopy (LIBS) in a Novel Molten Salt Aerosol System.

Science.gov (United States)

Williams, Ammon N; Phongikaroon, Supathorn

2017-04-01

In the pyrochemical separation of used nuclear fuel (UNF), fission product, rare earth, and actinide chlorides accumulate in the molten salt electrolyte over time. Measuring this salt composition in near real-time is advantageous for operational efficiency, material accountability, and nuclear safeguards. Laser-induced breakdown spectroscopy (LIBS) has been proposed and demonstrated as a potential analytical approach for molten LiCl-KCl salts. However, all the studies conducted to date have used a static surface approach which can lead to issues with splashing, low repeatability, and poor sample homogeneity. In this initial study, a novel molten salt aerosol approach has been developed and explored to measure the composition of the salt via LIBS. The functionality of the system has been demonstrated as well as a basic optimization of the laser energy and nebulizer gas pressure used. Initial results have shown that this molten salt aerosol-LIBS system has a great potential as an analytical technique for measuring the molten salt electrolyte used in this UNF reprocessing technology.

Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel; Estudio de sistema de un proceso de tratamiento-reciclaje piroquimico del combustible de un reactor de sales fundidas

Energy Technology Data Exchange (ETDEWEB)

Boussier, H.; Heuer, D.

2010-07-01

The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Fast Reactor (MSFR).

Fuel processing for molten-salt reactors

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1976-01-01

Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, April-June 1978. [Advanced solvent extraction; accidents; pyrochemical; radwaste in metal matrix; waste migration

Energy Technology Data Exchange (ETDEWEB)

Steindler, M. J.; Ader, M.; Barletta, R. E.

1979-12-01

Fuel cycle studies reported include development of centrifugal contactors for Purex processes. Tricaprylmethyl-ammonium nitrate and di-n-amyl-n-amylphosphonate are being evaluated as Thorex extractants. Dispersion of uranium and plutonium by fires, and mechanisms for subdividing and dispersing liquids and solids were reviewed. In the pyrochemical and dry processing program, a facility for testing containment materials is under construction; a flowsheet for carbide fuel processing has been designed and studies of carbide reactions in bismuth are underway; salt transport processes are being studied; process-size refractory metal vessels are being fabricated; the feasibility of AIROX reprocessing is being determined; the solubility of UO/sub 2/, UO/sub 2/ + fission products, and PuO/sub 2/ in molten alkali metal nitrates, has been investigated; a flowsheet was developed for reprocessing actinide oxides in molten salts; preparation of Th-U carbide from the oxide is being studied; new flowsheets based on the Dow Aluminum Pyrometallurgical process for reprocessing of spent uranium metal fuel have been prepared; the chloride volitility processing of thorium-based fuels is being studied; the reprocessing of (Th,U)O/sub 2/ solid solution in KCl-LiCl-ThCl/sub 4/-Th is being studied; and a flowsheet for processing spent nuclear fuel in molten tin has been constructed. Leach rates of simulated encapsulated waste forms in a metal matrix were studied. Nine criteria for handling waste cladding hulls were established. Strontium and tin migration in glauconite columns was measured. Radioactive Sr in a stream of water moved through oolitic limestone as rapidly as water, but in a stream of water equilibrated with the limestone, Sr moved through the limestone one-tenth as fast. Migration of trace quantities of Cs and I through kaolinite was studied. 88 figures, 53 tables.

Electrochemical reduction of cerium oxides in molten salts

International Nuclear Information System (INIS)

Claux, B.; Serp, J.; Fouletier, J.

2011-01-01

This brief article describes a pyrochemical process that is used by CEA to turn actinide oxides into metal actinides. This process is applied to Cerium oxides (CeO 2 ) that simulate actinide oxides well chemically as cerium belongs to the lanthanide family. The process is in fact an electrolysis of cerium oxide in a bath of molten calcium chloride salt whose temperature is between 800 and 900 Celsius degrees. At those temperatures calcium chloride becomes a ionic liquid (Ca 2+ and Cl - ) that is a good electrical conductor and is particularly well-adapted as solvent to an electrolytic process. The electrolysis current allows the transformation of solvent Ca 2+ ions into metal calcium which, in turn, can reduce cerium oxide into metal cerium through chromatically. Experimental data shows the reduction of up to 90% of 10 g samples of CeO 2 in a 6 hour long electrolysis while the best reduction rate ever known was 80% so far. This result is all the more promising that cerium oxides are more difficult to reduce than actinide oxides from the thermodynamical perspective

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

Pyrochemical processes for the recovery of weapons grade plutonium either as a metal or as PuO2 for use in mixed oxide reactor fuel pellets

International Nuclear Information System (INIS)

Colmenares, C.A.; Ebbinghaus, B.B.; Bronson, M.C.

1995-01-01

The authors have developed two processes for the recovery of weapons grade Pu, as either Pu metal or PuO 2 , that are strictly pyrochemical and do not produce any liquid waste. Large amounts of Pu metal (up to 4 kg.), in various geometric shapes, have been recovered by a hydride/dehydride/casting process (HYDEC) to produce metal ingots of any desired shape. The three processing steps are carried out in a single compact apparatus. The experimental technique and results obtained will be described. The authors have prepared PuO 2 powders from weapons grade Pu by a process that hydrides the Pu metal followed by the oxidation of the hydride (HYDOX process). Experimental details of the best way to carry out this process will be presented, as well as the characterization of both hydride and oxide powders produced

Study Effect of Salt Washing Process on Content and Iodium Stability of Salt

Directory of Open Access Journals (Sweden)

Nelson Saksono

2010-10-01

Full Text Available Effect of Salt Washing Process on Content and Iodium Stability of Salt. Salt washing process should increase the saltquality. It should clean the salt from sludge or clay and also reduce the impurity compound such as Mg, Ca and the reductor content. The objective of these reseach is to assess the effect of washing process on the content og hygroscopic impurities compound (Ca and Mg, and reductor content of salt. The research also investigate the water absorbing, pH, KIO3 content as function of time to obtain effect of washing process on KIO3 stability in salt. The experiment result shows that the lowest content of Mg and reductor compound 0.016 % wt and 2.65 ppm respectively which is reached at the fi ne salt washing process using 27 % wt brine. The analysis of water content indicates an increase the Ca and Mg content, causing an water absorbtion in salt , However the effect on pH the is not clear.

Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

International Nuclear Information System (INIS)

Usami, T.; Kurata, M.; Inoue, T.; Sims, H.E.; Beetham, S.A.; Jenkins, J.A.

2002-01-01

The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li 2 O in LiCl was measured to be 8.8 wt% at 650 deg. C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO 2 to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%

Solubility of uranium in liquid gallium, indium and their alloys

International Nuclear Information System (INIS)

Volkovich, Vladimir A.; Maltsev, Dmitry S.; Yamschikov, Leonid F.; Osipenko, Alexander G.; Kormilitsyn, Mikhail V.

2014-01-01

Pyrochemical reprocessing of spent nuclear fuels (SNF) employing molten salts and liquid metals as working media is considered as a possible alternative to the existing liquid extraction (PUREX) processes. Liquid salts and metals allow reprocessing highly irradiated high burn-up fuels with short cooling times, including the fuels of fast neutron reactors. Pyrochemical technology opens a way to practical realization of short closed fuel cycle. Liquid low-melting metals are immiscible with molten salts and can be effectively used for separation (or selective extraction) of SNF components dissolved in fused salts. Binary or ternary alloys of eutectic compositions can be employed to lower the melting point of the metallic phase. However, the information on SNF components behaviour and properties in ternary liquid metal alloys is very scarce

Lithium actinide recycle process demonstration

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.K.; Pierce, R.D.; McPheeters, C.C. [Argonne National Laboratory, IL (United States)

1995-10-01

Several pyrochemical processes have been developed in the Chemical Technology Division of Argonne Laboratory for recovery of actinide elements from LWR spent fuel. The lithium process was selected as the reference process from among the options. In this process the LWR oxide spent fuel is reduced by lithium at 650{degrees}C in the presence of molten LiCl. The Li{sub 2}O formed during the reduction process is soluble in the salt. The spent salt and lithium are recycled after the Li{sub 2}O is electrochemically reduced. The oxygen is liberated as CO{sub 2} at a carbon anode or oxygen at an inert anode. The reduced metal components of the LWR spent fuel are separated from the LiCL salt phase and introduced into an electrorefiner. The electrorefining step separates the uranium and transuranium (TRU) elements into two product streams. The uranium product, which comprises about 96% of the LWR spent fuel mass, may be enriched for recycle into the LWR fuel cycle, stored for future use in breeder reactors, or converted to a suitable form for disposal as waste. The TRU product can be recycled as fast reactor fuel or can be alloyed with constituents of the LWR cladding material to produce a stable waste form.

Estimation of centerline temperature of the waste form for the rare earth waste generated from pyrochemical process

Energy Technology Data Exchange (ETDEWEB)

Choi, Jung-Hoon, E-mail: mrchoijh@kaeri.re.kr; Eun, Hee-Chul; Lee, Tae-Kyo; Lee, Ki-Rak; Han, Seung-Youb; Jeon, Min-Ku; Park, Hwan-Seo; Ahn, Do-Hee

2017-01-15

Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO{sub 2}−Al{sub 2}O{sub 3}−B{sub 2}O{sub 3} glass frit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form.

Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

International Nuclear Information System (INIS)

Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

1998-10-01

Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

Evaluation of Li3N accumulation in a fused LiCl/Li salt matrix

International Nuclear Information System (INIS)

Eberle, C. S.

1998-01-01

Pyrochemical conditioning of spent nuclear fuel for the purpose of final disposal is currently being demonstrated at Argonne National Laboratory (ANL), and ongoing research in this area includes the demonstration of this process on spent oxide fuel. In conjunction with this research a pilot scale of the preprocessing stage is being designed by ANL-W to demonstrate the in situ hot cell capability of the chemical reduction stage. An impurity evaluation was completed for a Li/LiCl salt matrix in the presence of spent LWR uranium oxide fuel. A simple analysis was performed in which the sources of impurities in the salt matrix were only from the cell atmosphere. Only reactions with the lithium were considered. The levels of impurities were shown to be highly sensitive system conditions. A predominance diagram for the Li-O-N system was constructed for the device, and the general oxidation, nitridation and combined reactions were calculated as a function of oxygen and nitrogen partial pressure. These calculations and hotcell atmosphere data were used to determine the total number and type of impurities expected in the salt matrix and the mass rate for the device was determined

Application of lithium in molten-salt reduction processes

International Nuclear Information System (INIS)

Gourishankar, K. V.

1998-01-01

Metallothermic reductions have been extensively studied in the field of extractive metallurgy. At Argonne National Laboratory (ANL), we have developed a molten-salt based reduction process using lithium. This process was originally developed to reduce actinide oxides present in spent nuclear fuel. Preliminary thermodynamic considerations indicate that this process has the potential to be adapted for the extraction of other metals. The reduction is carried out at 650 C in a molten-salt (LiCl) medium. Lithium oxide (Li 2 O), produced during the reduction of the actinide oxides, dissolves in the molten salt. At the end of the reduction step, the lithium is regenerated from the salt by an electrowinning process. The lithium and the salt from the electrowinning are then reused for reduction of the next batch of oxide fuel. The process cycle has been successfully demonstrated on an engineering scale in a specially designed pyroprocessing facility. This paper discusses the applicability of lithium in molten-salt reduction processes with specific reference to our process. Results are presented from our work on actinide oxides to highlight the role of lithium and its effect on process variables in these molten-salt based reduction processes

Characteristics of dechlorination for LiCl salt by the surface temperature-controlled reactor

Energy Technology Data Exchange (ETDEWEB)

Cho, In Hak [Chungnam National University, Daejeon (Korea, Republic of); Park, Hwan Seo; Ahn, Soo Na; Eun, Hee Chul; Kim, In Tae; Cho, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Molten salt waste is generated from a pyrochemical process to separate reusable U and TRU elements from a spent nuclear fuel. The spent lithium chloride waste is highly soluble in water and contains volatile radioactive elements such as Cs. However, these wastes are difficult to directly immobilize into durable matrix such as glass or ceramic wasteform for final disposal. ANL(Argonne National Laboratory) suggested the conversion of metal chloride into a sodalite for the immobilization of a chloride waste, glass-bonded sodalite, which was fabricated at about 915 .deg. C after mixing the salt-loaded zeolite and borosilicate glass powder. Although this wasteform shows high leach-resistance, the waste volume greatly increases. The previous study was to treat metal chloride wastes by using SAP(SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) materials. By using this method, the final waste volume reduced and leach-resistance was good. In this study, characteristics of dechlorination reaction of LiCl with an inorganic composite, SAP, was investigated by using a specific surface temperature-controlled reactor

Deployment of quasi-digital sensor for high temperature molten salt level measurement in pyroprocessing plants

Science.gov (United States)

Sanga, Ramesh; Agarwal, Sourabh; Sivaramakrishna, M.; Rao, G. Prabhakara

2018-04-01

Development of a liquid molten salt level sensor device that can detect the level of liquid molten salt in the process vessels of pyrochemical reprocessing of spent metallic fuels is detailed. It is proposed to apply a resistive-type pulsating sensor-based level measurement approach. There are no commercially available sensors due to limitations of high temperature, radiation, and physical dimensions. A compact, simple, rugged, low power, and high precise pulsating sensor-based level probe and simple instrumentation for the molten salt liquid level sensor to work in the extreme conditions has been indigenously developed, with high precision and accuracy. The working principle, design concept, and results have been discussed. This level probe is mainly composed of the variable resistor made up of ceramic rods. This resistor constitutes the part of resistance-capacitance-type Logic Gate Oscillator (LGO). A change in the molten salt level inside the tank causes a small change in the resistance which in turn changes the pulse frequency of the LGO. Thus the frequency, the output of the instrument that is displayed on the LCD of an embedded system, is a function of molten salt level. In the present design, the range of level measurement is about 10 mm. The sensitivity in position measurement up to 10 mm is ˜2.5 kHz/mm.

Applications of molten salts in plutonium processing

International Nuclear Information System (INIS)

Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

1987-01-01

Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900 0 C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics

Simulated real-time process monitoring of a molten salt electrorefiner

International Nuclear Information System (INIS)

Rappleye, Devin; Simpson, Michael; Cumberland, Riley; McNelis, David; Yim, Man-Sung

2014-01-01

Highlights: • An alternative approach to safeguarding and monitoring pyroprocessing is proposed. • Possible signals to be used to monitor an electrorefiner are identified. • An inverse model was developed to determine deposition rates at the cathode. • The sensitivity of certain parameters in the inverse model are presented. - Abstract: An alternative approach to monitoring the pyrochemical process (pyroprocessing) for spent nuclear fuel treatment is proposed and examined. This approach relies on modeling and the real-time analysis of process readings. Using an electrorefiner model, named ERAD, cathode potential and cell current were identified as useful process readings. To provide a real-time analysis of these two process readings, an inverse model was developed based on fundamental electrochemical relations. The model was applied to the following operating modes: pure uranium deposition, co-deposition of uranium and plutonium, and co-deposition of uranium and zirconium. Using the cell current and cathode potential, the model predicted which species were depositing and their rates. The deposition rates predicted by the inverse model compared favorably to those calculated by ERAD

De-chlorination and solidification of radioactive LiCl waste salt by using SiO_2-Al_2O_3-P_2O_5 (SAP) inorganic composite including B_2O_3 component

International Nuclear Information System (INIS)

Lee, Ki Rak; Park, Hwan-Seo; Cho, In-Hak; Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Han, Seung Youb; Ahn, Do-Hee

2017-01-01

SAP (SiO_2-Al_2O_3-P_2O_5) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.

Molten salt data for pyro chemistry: new research advanced and development of information system

International Nuclear Information System (INIS)

Gaune-Escard, M.; Adya, A.K.; Rycerz, L.

2001-01-01

Pyrochemical reprocessing of nuclear fuels simultaneously involves materials, energy production and environmental aspects. However, the data relative to molten lanthanide and actinide salts (and compounds), specific to this field, are not easily available. Because of characteristic physicochemical properties, which do not promote experimental investigations, data are scarce. Very often, most of the data is buried in grey literature. When available, they reveal to arise from estimations. Co-ordination and standardisation of existing data is therefore essential, as research efforts should be intensified together with development of numerical prediction tools. This paper aims, firstly, to discuss the new experimental data obtained through an intensive international research co-operation and, secondly, to present ongoing efforts for the development of a molten salt database. (author)

Molten salt processes in special materials preparation

International Nuclear Information System (INIS)

Krishnamurthy, N.; Suri, A.K.

2013-01-01

As a class, molten salts are the largest collection of non aqueous inorganic solvents. On account of their stability at high temperature and compatibility to a number of process requirements, molten salts are considered indispensable to realize many of the numerous benefits of high temperature technology. They play a crucial role and form the basis for numerous elegant processes for the preparation of metals and materials. Molten salt are considered versatile heat transfer media and have led to the evolution of many interesting reactor concepts in fission and possibly in fusion. They also have been the basis of thinking for few novel processes for power generation. While focusing principally on the actual utilization of molten salts for a variety of materials preparation efforts in BARC, this lecture also covers a few of the other areas of technological applications together with the scientific basis for considering the molten salts in such situations. (author)

Simulation of salt production process

Science.gov (United States)

Muraveva, E. A.

2017-10-01

In this paper an approach to the use of simulation software iThink to simulate the salt production system has been proposed. The dynamic processes of the original system are substituted by processes simulated in the abstract model, but in compliance with the basic rules of the original system, which allows one to accelerate and reduce the cost of the research. As a result, a stable workable simulation model was obtained that can display the rate of the salt exhaustion and many other parameters which are important for business planning.

Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

Energy Technology Data Exchange (ETDEWEB)

Kim Davies; Shelly X Li

2007-09-01

Pyrochemical processing plays an important role in development of proliferation- resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ~2 grams of LiCl/KCl salt electrolyte with a low concentration (~1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver wire

Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

International Nuclear Information System (INIS)

Kim Davies; Shelly X Li

2007-01-01

Pyrochemical processing plays an important role in development of proliferation-resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ∼2 grams of LiCl/KCl salt electrolyte with a low concentration (∼1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver

Dehydration of ethanol with salt extractive distillation-a comparative analysis between processes with salt recovery

Energy Technology Data Exchange (ETDEWEB)

Ligero, E.L.; Ravagnani, T.M.K. [Departamento de Engenharia de Sistemas Qumicos, Faculdade de Engenharia Qumica, Universidade Estadual de Campinas, Campinas, Sao Paulo (Brazil)

2003-07-01

Anhydrous ethanol can be obtained from a dilute aqueous solution of ethanol via extractive distillation with potassium acetate. Two process flowsheets with salt recovery were proposed. In the first, dilute ethanol is directly fed to a salt extractive distillation column and, after that, the salt is recovered in a multiple effect evaporator followed by a spray dryer. In the second, the concentrated ethanol from conventional distillation is fed to a salt extractive distillation column. In this case, salt is recovered in a single spray dryer. In both processes the recovered salt is recycled to be used in the extractive distillation column. Every component of each process was rigorously modeled and its behavior was simulated for a wide range of operating conditions. A global simulation was then carried out. The results show that the second process is more interesting in terms of energy consumption than the first. Furthermore, it would be easier to implement changes on existing benzene extractive anhydrous ethanol plants to convert them to more ecologically attractive concentrated ethanol feed processes. (author)

MSO spent salt clean-up recovery process; TOPICAL

International Nuclear Information System (INIS)

Adamson, M G; Brummond, W A; Hipple, D L; Hsu, P C; Summers, L J; Von Holtz, E H; Wang, F T

1997-01-01

An effective process has been developed to separate metals, mineral residues, and radionuclides from spent salt, a secondary waste generated by Molten Salt Oxidation (MSO). This process includes salt dissolution, pH adjustment, chemical reduction and/or sulfiding, filtration, ion exchange, and drying. The process uses dithionite to reduce soluble chromate and/or sulfiding agent to suppress solubilities of metal compounds in water. This process is capable of reducing the secondary waste to less than 5% of its original weight. It is a low temperature, aqueous process and has been demonstrated in the laboratory[1

Electrochemical studies on plutonium in molten salts

International Nuclear Information System (INIS)

Bourges, G.; Lambertin, D.; Rochefort, S.; Delpech, S.; Picard, G.

2007-01-01

Electrochemical studies on plutonium have been supporting the development of pyrochemical processes involving plutonium at CEA. The electrochemical properties of plutonium have been studied in molten salts - ternary eutectic mixture NaCl-KCl-BaCl 2 , equimolar mixture NaCl-KCl and pure CaCl 2 - and in liquid gallium at 1073 K. The formal, or apparent, standard potential of Pu(III)/Pu redox couple in eutectic mixture of NaCl-KCl-BaCl 2 at 1073 K determined by potentiometry is equal to -2.56 V (versus Cl 2 , 1 atm/Cl - reference electrode). In NaCl-KCl eutectic mixture and in pure CaCl 2 the formal standard potentials deduced from cyclic voltammetry are respectively -2.54 V and -2.51 V. These potentials led to the calculation of the activity coefficients of Pu(III) in the molten salts. Chronoamperometry on plutonium in liquid gallium using molten chlorides - CaCl 2 and equimolar NaCl/KCl - led to the determination of the activity coefficient of Pu in liquid Ga, log γ = -7.3. This new data is a key parameter to assess the thermodynamic feasibility of a process using gallium as solvent metal. By comparing gallium with other solvent metals - cadmium, bismuth, aluminum - gallium appears to be, with aluminum, more favorable for the selectivity of the separation at 1073 K of plutonium from cerium. In fact, compared with a solid tungsten electrode, none of these solvent liquid metals is a real asset for the selectivity of the separation. The role of a solvent liquid metal is mainly to trap the elements

ADR salt pill design and crystal growth process for hydrated magnetic salts

Science.gov (United States)

Shirron, Peter J. (Inventor); DiPirro, Michael J. (Inventor); Canavan, Edgar R. (Inventor)

2013-01-01

A process is provided for producing a salt pill for use in very low temperature adiabatic demagnetization refrigerators (ADRs). The method can include providing a thermal bus in a housing. The thermal bus can include an array of thermally conductive metal conductors. A hydrated salt can be grown on the array of thermally conductive metal conductors. Thermal conductance can be provided to the hydrated salt.

Processing of effluent salt from the direct oxide reduction process

International Nuclear Information System (INIS)

Mishra, B.; Olson, D.L.

1992-01-01

The production of reactive metals by Direct Oxide Reduction (DOR) process using calcium in a molten calcium salt system generates significant amount of contaminated waste as calcium oxide saturated calcium chloride salt mix with calcium oxide content of up to 15 wt. pct. Fused salt electrolysis of a simulated salt mix has been carried out to electrowin calcium, which can be recycled to the DOR reactor along with the calcium chloride salt or may be used in-situ in a combined DOR and electrowinning process. Many reactive metal oxides could thus be reduced in a one-step process without generating a significant amount of waste. The process has been optimized in terms of the calcium solubility, cell temperature, current density and the cell design to maximize the current efficiency. Based on the information available regarding the solubility of calcium in calcium chloride salt in the presence of calcium oxide, and the back reactions occurring in-situ between the electrowon calcium and other components present in the cell, e.g. carbon, oxygen, carbon dioxide and calcium oxide, it is difficult to recover elemental calcium within the system. However, a liquid cathode or a rising cathode has been used in the past to recover calcium. The solubility has also been found to depend on the use of graphite as the anode material as evidenced by the presence of calcium carbonate in the final salt. The rate of recovery for metallic calcium has to be enhanced to levels that overcome the back reactions in a system where quick removal of anodic gases is achieved. Calcium has been detected by the hydrogen evolution technique and the amount of calcia has been determined by titration. A porous ceramic sheath has been used in the cell to prevent the chemical reaction of electrowon calcium to produce oxide or carbonate and to prevent the contamination of salt by the anodic carbon

Recovery of plutonium by pyroredox processing

International Nuclear Information System (INIS)

McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

1985-09-01

Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 7 figs., 4 tabs

Recovery of plutonium by pyroredox processing

International Nuclear Information System (INIS)

McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

1985-01-01

Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 2 figs., 5 tabs

De-chlorination and solidification of radioactive LiCl waste salt by using SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5} (SAP) inorganic composite including B{sub 2}O{sub 3} component

Energy Technology Data Exchange (ETDEWEB)

Lee, Ki Rak; Park, Hwan-Seo; Cho, In-Hak; Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Han, Seung Youb; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-09-15

SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.

Pyrochemical recovery of plutonium from calcium fluoride reduction slag

Science.gov (United States)

Christensen, D.C.

A pyrochemical method of recovering finely dispersed plutonium metal from calcium fluoride reduction slag is claimed. The plutonium-bearing slag is crushed and melted in the presence of at least an equimolar amount of calcium chloride and a few percent metallic calcium. The calcium chloride reduces the melting point and thereby decreases the viscosity of the molten mixture. The calcium reduces any oxidized plutonium in the mixture and also causes the dispersed plutonium metal to coalesce and settle out as a separate metallic phase at the bottom of the reaction vessel. Upon cooling the mixture to room temperature, the solid plutonium can be cleanly separated from the overlying solid slag, with an average recovery yield on the order of 96 percent.

Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

Directory of Open Access Journals (Sweden)

Rodrigues Davide

2015-12-01

Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

International Nuclear Information System (INIS)

Patterson-Hine, F.A.

1984-05-01

Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown

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

Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

Thorium nuclear fuel cycle technology

International Nuclear Information System (INIS)

Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

1998-03-01

Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

Sol-gel processing with inorganic metal salt precursors

Science.gov (United States)

Hu, Zhong-Cheng

2004-10-19

Methods for sol-gel processing that generally involve mixing together an inorganic metal salt, water, and a water miscible alcohol or other organic solvent, at room temperature with a macromolecular dispersant material, such as hydroxypropyl cellulose (HPC) added. The resulting homogenous solution is incubated at a desired temperature and time to result in a desired product. The methods enable production of high quality sols and gels at lower temperatures than standard methods. The methods enable production of nanosize sols from inorganic metal salts. The methods offer sol-gel processing from inorganic metal salts.

Salt processed food and gastric cancer in a Chinese population.

Science.gov (United States)

Lin, Si-Hao; Li, Yuan-Hang; Leung, Kayee; Huang, Cheng-Yu; Wang, Xiao-Rong

2014-01-01

To investigate the association between salt processed food and gastric cancer, a hospital based case-control study was conducted in a high risk area of China. One hundred and seven newly diagnosed cases with histological confirmation of gastric cancer and 209 controls were recruited. Information on dietary intake was collected with a validated food frequency questionnaire. Unconditional logistic regression was applied to estimate the odds ratios with adjustment for other potential confounders. Comparing the high intake group with never consumption of salt processed foods, salted meat, pickled vegetables and preserved vegetables were significantly associated with increased risk of gastric cancer. Meanwhile, salt taste preference in diet showed a dose-response relationship with gastric cancer. Our results suggest that consumption of salted meat, pickled and preserved vegetables, are positively associated with gastric cancer. Reduction of salt and salt processed food in diets might be one practical measure to preventing gastric cancer.

[Food processing industry--the salt shock to the consumers].

Science.gov (United States)

Doko Jelinić, Jagoda; Nola, Iskra Alexandra; Andabaka, Damir

2010-05-01

Industrial food production and processing is necessarily connected with the use of salt. Salt or sodium chloride is used as a preservative, spice, agent for color maintenance, texture, and to regulate fermentation by stopping the growth of bacteria, yeast and mold. Besides kitchen salt, other types of salt that also contain sodium are used in various technological processes in food preparing industry. Most of the "hidden" salt, 70%-75%, can be brought to the body by using industrial food, which, unfortunately, has been increasingly used due to the modern way of life. Bread and bakery products, meat products, various sauces, dried fish, various types of cheese, fast food, conserved vegetables, ready-made soups and food additives are the most common industrial foods rich in sodium. Many actions have been taken all over the world to restrict salt consumption. The World Health Organization recommends the upper limit of salt input of 5 g per day. These actions appeal to food industry to reduce the proportion of salt in their products. Besides lower salt addition during manufacture, food industry can use salt substitutes, in particular potassium chloride (KCl), in combination with additives that can mask the absence of salt, and flavor intensifiers that also enhance the product salinity. However, food industry is still quite resistant to reducing salt in their products for fear from losing profits.

Assessment of crushed salt consolidation and fracture healing processes in a nuclear waste repository in salt

International Nuclear Information System (INIS)

1984-11-01

For a nuclear waste repository in salt, two aspects of salt behavior are expected to contribute to favorable conditions for waste isolation. First, consolidation of crushed salt backfill due to creep closure of the underground openings may result in a backfill barrier with low permeability. Second, fractures created in the salt by excavation may heal under the influence of stress and temperature following sealing. This report reviews the status of knowledge regarding crushed salt consolidation and fracture healing, provides analyses which predict the rates at which the processes will occur under repository conditions, and develops requirements for future study. Analyses of the rate at which crushed salt will consolidate are found to be uncertain because of unexplained wide variation in the creep properties of crushed salt obtained from laboratory testing, and because of uncertainties in predictions of long term closure rates of openings in salt. This uncertainty could be resolved to a large degree by additional laboratory testing of crushed salt. Similarly, additional testing of fracture healing processes is required to confirm that healing will be effective under repository conditions. Extensive references, 27 figures, 5 tables

Treatment of plutonium process residues by molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)

1999-04-01

Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

Treatment of plutonium process residues by molten salt oxidation

International Nuclear Information System (INIS)

Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.

1999-01-01

Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 238 Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na 2 SO 4 , Na 3 PO 4 and NaAsO 2 or Na 3 AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the 238 Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox

Molten salt processing of mixed wastes with offgas condensation

International Nuclear Information System (INIS)

Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

1991-01-01

We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

Preview of DATACHOA Ver. 1.0: Database for Chemical and Physical Properties of Molten Chloride Salts Comprising Actinide and Lanthanide Chloride Salts

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong-Yun; Lee, Jin-Hong; Park, Tae-Hong; Bae, Sang-Eun; Cho, Young Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

International Atomic Energy Agency (IAEA) provides molten salt database based on a world-wide-web (WWW) server. However, many data useful for the pyroprocess under development in Korea are missing. We have developed various measurement techniques for obtaining the chemical and physical data of the molten salts over the last four years: absorption spectra, emission spectra, molar absorptivity, Gibbs free energy, apparent potential, activity coefficient, exchange current density, solubility, electrical conductivity, density, surface tension, viscosity, melting point, diffusion coefficient, etc, which have been of our current interests. Those property data are essential in the understanding and the operation of the pyrochemical system. Some properties are intercorrelated, and multi-component molten salt systems can be understood by using a multivariate chemometric data analysis. Our final goal is to establish the 'Expert System', which is defined as a system that emulates the decision-making ability of a human expert. Although expert systems have been successful in the artificial intelligence, they are not known to chemists or chemical engineers for many years. The DATACHOA is the first step forward to the 'Expert System', although the amount of our property data is not currently enough to build up the 'Expert System'. DATACHOA is our first attempt to collect extensive basic and fundamental properties of molten salts of interest to pyroprocess chemists and engineers in Korea. DATACHOA ver. 1.0 will be released to the pyroprocess engineers within a few months.

Development of fluoride reprocessing technology for molten salt transmutation reactor systems in the Czech Republic

International Nuclear Information System (INIS)

Uhlir, J.; Hosnedl, P.; Matal, O.

2000-01-01

At present, the transmutation of spent nuclear fuel is considered a prospective alternative conception with respect to the current conception based on the non-reprocessed spent fuel disposal into a deep geological repository. The Czech research and development programme in the area of partitioning is directed primarily on the development of the fuel cycle technology for the accelerator - driven subcritical reactor with a liquid fuel based on fluoride melts. The final objective of the research programme is the development of pyrochemical technologies suitable for a continuous or semi-continuous separation process which would allow practically perfect utilization of the transmutation potentialities of the reactor system. The present research is directed particularly on the development of suitable fluoride separation methods the target of which is the removal of the uranium component from spent nuclear fuel and on the research of the electro-separation procedures and further on the development of appropriate construction materials and equipment for the technology of fluoride salt melts. (authors)

Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

International Nuclear Information System (INIS)

Patterson-Hine, F.A.; Davidson, J.W.; Klein, D.E.; Lee, J.D.

1985-01-01

Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% 233 U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% 233 U case and 5225 kg/year for the 1.0% 233 U case

Plutonium metal preparation and purification at Los Alamos, 1984

International Nuclear Information System (INIS)

Christensen, D.C.; Williams, J.D.; McNeese, J.A.; Fife, K.W.

1984-01-01

Plutonium metal preparation and purification are well established at Los Alamos. Metal is prepared by calcothermic reduction of both PuF 4 and PuO 2 . Metal is purified by halide slagging, casting, and electrorefining. The product from the production sequence is ultrapure plutonium metal. All of the processes involve high temperature operation and all but casting involve molten salt media. Development efforts are fourfold: (1) recover plutonium values from residues; (2) reduce residue generation through process improvements and changes; (3) recycle of reagents, and (4) optimize and integrate all processes into a close-loop system. Plutonium residues are comprised of oxides, chlorides, colloidal metal suspensions, and impure metal heels. Pyrochemical recovery techniques are under development to address each residue. In addition, we are looking back at each residue generation step and are making process changes to reduce plutonium content in each residue. Reagent salt is the principle media used in pyrochemical processing. The regeneration and recycle of these reagents will both reduce our waste handling and operating expense. The fourth area, process optimization, involves both existing processes and new process developments. A status of efforts in all four of these areas will be summarized

Process for the preparation of protected dihydroxypropyl trialkylammonium salts and derivatives thereof

Energy Technology Data Exchange (ETDEWEB)

Hollingsworth, R.I.; Wang, G.

2000-07-04

A process for the preparation of protected dihydroxypropyl trialkylammonium salts, particularly in chiral form is described. In particular, a process for the preparation of (2,2-dimethyl-1,3-dioxolan-4-ylmethyl)trialkylammonium salts, particularly in chiral form is described. Furthermore, a process is described wherein the (2,2-dimethyl-1,3-dioxolan-4ylmethyl)trialkylammonium salts is a 2,2-dimethyl-1,3-dioxolan-4-ylmethyl trimethylammonium salt, preferably in chiral form. The protected dihydroxypropyl trialkylammonium salts lead to L-carnitine when in chiral form.

Process for the preparation of protected dihydroxypropyl trialkylammonium salts and derivatives thereof

Energy Technology Data Exchange (ETDEWEB)

Hollingsworth, Rawle I. (Haslett, MI); Wang, Guijun (East Lansing, MI)

2000-01-01

A process for the preparation of protected dihydroxypropyl trialkylammonium salts, particularly in chiral form is described. In particular, a process for the preparation of (2,2-dimethyl-1,3-dioxolan-4-ylmethyl)trialkylammonium salts, particularly in chiral form is described. Furthermore, a process is described wherein the (2,2-dimethyl-1,3-dioxolan-4ylmethyl)trialkylammonium salts is a 2,2-dimethyl-1,3-dioxolan-4-ylmethyl trimethylammonium salt, preferably in chiral form. The protected dihydroxypropyl trialkylammonium salts lead to L-carnitine (9) when in chiral form (5).

Organic waste processing using molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M. G., LLNL

1998-03-01

Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

Removal of uranium from spent salt from the moltensalt oxidation process

International Nuclear Information System (INIS)

Summers, L.; Hsu, P.C.; Holtz, E.V.; Hipple, D.; Wang, F.; Adamson, M.

1997-03-01

Molten salt oxidation (MSO) is a thermal process that has the capability of destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials. In this process, combustible waste and air are introduced into the molten sodium carbonate salt. The organic constituents of the waste materials are oxidized to carbon dioxide and water, while most of the inorganic constituents, including toxic metals, minerals, and radioisotopes, are retained in the molten salt bath. As these impurities accumulate in the salt, the process efficiency drops and the salt must be replaced. An efficient process is needed to separate these toxic metals, minerals, and radioisotopes from the spent carbonate to avoid generating a large volume of secondary waste. Toxic metals such as cadmium, chromium, lead, and zinc etc. are removed by a method described elsewhere. This paper describes a separation strategy developed for radioisotope removal from the mixed spent salt, as well as experimental results, as part of the spent salt cleanup. As the MSO system operates, inorganic products resulting from the reaction of halides, sulfides, phosphates, metals and radionuclides with carbonate accumulate in the salt bath. These must be removed to prevent complete conversion of the sodium carbonate, which would result in eventual losses of destruction efficiency and acid scrubbing capability. There are two operational modes for salt removal: (1) during reactor operation a slip-stream of molten salt is continuously withdrawn with continuous replacement by carbonate, or (2) the spent salt melt is discharged completely and the reactor then refilled with carbonate in batch mode. Because many of the metals and/or radionuclides captured in the salt are hazardous and/or radioactive, spent salt removed from the reactor would create a large secondary waste stream without further treatment. A spent salt clean up/recovery system is necessary to segregate these materials and minimize the amount of

The electrodeposition and rare earths reduction in the molten salt actinides recovery systems using liquid metal

International Nuclear Information System (INIS)

Shim, J-B.; Lee, J-H.; Kwon, S-W.; Ahn, B-G.; Woo, M-S.; Lee, B-J.; Kim, E-H.; Park, H-S.; Yoo, J-H.

2005-01-01

A pyrochemical partitioning system uses liquid metals such as cadmium and bismuth in order to recover the actinide metals from a molten salt mixture containing rare earth fission product metals. The liquid metals play roles as a cathode in the electrowinning or an extracting phase in the reductive extraction operation. The product resulting from the above operations is metal-cadmium or-bismuth alloy, which should contain the rare earth element amounts as low as possible for a transmutation purpose. In this study, the electrodeposition behaviours of uranium and lanthanide elements such as La, Ce and Nd were investigated for solid molybdenum and liquid cadmium electrodes in a molten LiCl-KCl eutectic salt. Electrochemical methods used are a cyclic voltammetry (CV) and a chronopotentiometry for monitoring the salt phase and recovering the metals, respectively. The CV graphs for monitoring the oxidizing agent CdCl 2 in the salt phase were obtained. These show a time dependently disappearance of the oxidizing agent corresponding to the formation of UCl 3 by inserting the uranium metal into the salt. Also, a sequential oxidation technique which is added at a controlled amount of the oxidizing agents into the salt phase was applied. It was found that this method is feasible for the selective reduction of the rare earths content in liquid metal alloys. (author)

Nitridation of U and Pu recovered in liquid Cd cathode by molten salt electrorefining of (U,Pu)N

Energy Technology Data Exchange (ETDEWEB)

Satoh, Takumi; Iwai, Takashi; Arai, Yasuo [Japan Atomic Energy Agency (Japan)

2009-06-15

Solid solutions of actinide mono-nitrides have been proposed as a candidate fuel of the accelerator-driven system (ADS) and Gen.IV-type fast reactors because the thermal conductivity and metal density are higher than those of actinide oxides and also they have high melting temperature. Pyrochemical process has several advantages over conventional wet process in treating of spent nitride fuel. One of the key technologies of the pyrochemical reprocessing of nitride fuel is the formation of the nitrides from actinides in the liquid Cd cathode. The nitridation-distillation combined method was developed and has been adopted to convert the actinides to the nitrides. In this method, the nitridation of actinides and the distillation of Cd occurred simultaneously by heating the actinide-Cd alloys in N{sub 2} gas stream. In the present study, the nitride formation behavior of U and Pu recovered in Cd cathode by molten salt electrorefining of (U,Pu)N was experimentally investigated. In addition, the nitride pellet was prepared form the powder obtained by the nitridation of U and Pu recovered in Cd cathode. (U,Pu)N (PuN = 80 mol %) was used as the starting material in the experiment. Molten salt electrorefining of (U,Pu)N pellet was carried out in the LiCl-KCl eutectic salt with 1.2 wt% PuCl{sub 3} and 0.3 wt% UCl{sub 3} of about 110 g at the constant anodic potential of -0.60 to -0.55 V vs. Ag/AgCl for about 9 hours at 773 K. After the electrorefining, about 42 % of U and Pu in the starting (U,Pu)N pellet was dissolved at the anode and recovered into the liquid Cd cathode. The recovered U-Pu-Cd alloy was heated in an alumina crucible at 973 K for 10 hours under N{sub 2} gas (99.999 %) stream (0.015 L/min). Fine black powder was recovered after heating the U-Pu-Cd alloy. The powder was identified as the single phase solid solution of (U,Pu)N by the XRD analysis. After milling in the agate mortar for 1 hour, the powder was compacted into green pellet under a pressure of about

Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

International Nuclear Information System (INIS)

Jain, V.; Shah, H.; Wilmarth, W. R.

2016-01-01

Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated

Modeling Coupled THM Processes and Brine Migration in Salt at High Temperatures

International Nuclear Information System (INIS)

Rutqvist, Jonny; Blanco-Martin, Laura; Molins, Sergi; Trebotich, David; Birkholzer, Jens

2015-01-01

In this report, we present FY2015 progress by Lawrence Berkeley National Laboratory (LBNL) related to modeling of coupled thermal-hydrological-mechanical-chemical (THMC) processes in salt and their effect on brine migration at high temperatures. This is a combined milestone report related to milestone Salt R&D Milestone ''Modeling Coupled THM Processes and Brine Migration in Salt at High Temperatures'' (M3FT-15LB0818012) and the Salt Field Testing Milestone (M3FT-15LB0819022) to support the overall objectives of the salt field test planning.

Modeling Coupled THM Processes and Brine Migration in Salt at High Temperatures

Energy Technology Data Exchange (ETDEWEB)

Rutqvist, Jonny [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Blanco-Martin, Laura [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Molins, Sergi [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Trebotich, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Birkholzer, Jens [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2015-09-01

In this report, we present FY2015 progress by Lawrence Berkeley National Laboratory (LBNL) related to modeling of coupled thermal-hydrological-mechanical-chemical (THMC) processes in salt and their effect on brine migration at high temperatures. This is a combined milestone report related to milestone Salt R&D Milestone “Modeling Coupled THM Processes and Brine Migration in Salt at High Temperatures” (M3FT-15LB0818012) and the Salt Field Testing Milestone (M3FT-15LB0819022) to support the overall objectives of the salt field test planning.

Recent development in pyrochemistry at Los Alamos

International Nuclear Information System (INIS)

McNeese, J.A.; Fife, K.W.; Williams, J.D.

1984-01-01

Recent developments in pyrochemical processing at Los Alamos include the recovery of plutonium from anodes and impure metal by pyroredox and new molten salt handling and purification techniques. The anode is dissolved in a ZnCl 2 KCl salt to form PuCl 3 and a zinc and impurities button. Calcium reduction of the PuCl 3 yields 95 to 98% pure plutonium. New techniques for transferring molten salt from a purification or regeneration vessel to molds has been successfully developed and demonstrated. Additional salt work involving recycle of direct oxide reduction salts using anhydrous hydrogen chloride, phosgene, and chlorine gases is under way. 13 figures, 1 table

Salt separation of uranium deposits generated from electrorefining in pyro process

International Nuclear Information System (INIS)

Kwon, S. W.; Park, K. M.; Jeong, J. H.; Lee, H. S.; Kim, J. G.

2012-01-01

Electrorefining is a key step in a pyro processing. Electrorefining process is generally composed of two recovery steps- deposit of uranium onto a solid cathode(electrorefining) and then the recovery of the remaining uranium and TRU(TransUranic) elements simultaneously by a liquid cadmium cathode(electrowinning). The uranium ingot is prepared from the deposits after the salt separation. In this study, the sequential operation of the liquid salt separation? distillation of the residual salt was attempted for the achievement of high throughput performance in the salt separation. The effects of deposit size and packing density were also investigated with steel chips, steel chips, and uranium dendrites. The apparent evaporation rate decreased with the increasing packing density or the increasing size of deposits due to the hindrance of the vapor transport by the deposits. It was found that the packing density and the geometry of deposit crucible are important design parameters for the salt separation system. Base on the results of the study, an engineering scale salt distiller was developed and installed in the argon cell. The salt distiller is a batch-type, and the process capacity to about 50 kg U-deposits/day. The design of the salt distiller is based on the remote operation by Master Slave Manipulator (MSM) and a hoist. The salt distiller is composed of two large blocks of the distillation tower and the crucible loading system for the transportation to maintenance room via the Large Transfer Lock (LTL)

Salt separation of uranium deposits generated from electrorefining in pyro process

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-15

Electrorefining is a key step in a pyro processing. Electrorefining process is generally composed of two recovery steps- deposit of uranium onto a solid cathode(electrorefining) and then the recovery of the remaining uranium and TRU(TransUranic) elements simultaneously by a liquid cadmium cathode(electrowinning). The uranium ingot is prepared from the deposits after the salt separation. In this study, the sequential operation of the liquid salt separation? distillation of the residual salt was attempted for the achievement of high throughput performance in the salt separation. The effects of deposit size and packing density were also investigated with steel chips, steel chips, and uranium dendrites. The apparent evaporation rate decreased with the increasing packing density or the increasing size of deposits due to the hindrance of the vapor transport by the deposits. It was found that the packing density and the geometry of deposit crucible are important design parameters for the salt separation system. Base on the results of the study, an engineering scale salt distiller was developed and installed in the argon cell. The salt distiller is a batch-type, and the process capacity to about 50 kg U-deposits/day. The design of the salt distiller is based on the remote operation by Master Slave Manipulator (MSM) and a hoist. The salt distiller is composed of two large blocks of the distillation tower and the crucible loading system for the transportation to maintenance room via the Large Transfer Lock (LTL)

Secondary Aluminum Processing Waste: Salt Cake Characterization and Reactivity

Science.gov (United States)

Thirty-nine salt cake samples were collected from 10 SAP facilities across the U.S. The facilities were identified by the Aluminum Association to cover a wide range of processes. Results suggest that while the percent metal leached from the salt cake was relatively low, the leac...

Nutritional modelling: distributions of salt intake from processed foods in New Zealand.

Science.gov (United States)

Thomson, Barbara M

2009-09-01

The salt content of processed foods is important because of the high intake of Na by most New Zealanders. A database of Na concentrations in fifty-eight processed foods was compiled from existing and new data and combined with 24 h diet recall data from two national nutrition surveys (5771 respondents) to derive salt intakes for seven population groups. Mean salt intakes from processed foods ranged from 6.9 g/d for young males aged 19-24 years to 3.5 g/d for children aged 5-6 years. A total of > or = 50 % of children aged 5-6 years, boys aged 11-14 years and young males aged 19-24 years had salt intakes that exceeded the upper limit for Na, calculated as salt (3.2-5.3 g/d), from processed foods only. Bread accounted for the greatest contribution to salt intake for each population group (35-43 % of total salt intake). Other foods that contributed 2 % or more and common across most age groups were sausage, meat pies, pizza, instant noodles and cheese. The Na concentrations of key foods have changed little over the 16-year period from 1987 to 2003 except for corned beef and whole milk that have decreased by 34 and 50 % respectively. Bread is an obvious target for salt reduction but the implication on iodine intake needs consideration as salt is used as a vehicle for iodine fortification of bread.

Combined gettering and molten salt process for tritium recovery from lithium

International Nuclear Information System (INIS)

Sze, D.K.; Finn, P.A.; Bartlit, J.; Tanaka, S.; Teria, T.; Yamawaki, M.

1988-02-01

A new tritium recovery concept from lithium has been developed as part of the US/Japan collaboration on Reversed-Field Pinch Reactor Design Studies. This concept combines the γ-gettering process as the front end to recover tritium from the coolant, and a molten salt recovery process to extract tritium for fuel processing. A secondary lithium is used to regenerate the tritium from the gettering bed and, in the process, increases the tritium concentration by a factor of about 20. That way, the required size of the molten salt process becomes very small. A potential problem is the possible poisoning of the gettering bed by the salt dissolved in lithium. 16 refs., 6 figs

Preconceptual design of a salt splitting process using ceramic membranes

Energy Technology Data Exchange (ETDEWEB)

Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R. [Pacific Northwest National Lab., Richland, WA (United States); Balagopal, S.; Landro, T.; Sutija, D.P. [Ceramatec, Inc., Salt Lake City, UT (United States)

1997-01-01

Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate.

Preconceptual design of a salt splitting process using ceramic membranes

International Nuclear Information System (INIS)

Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R.; Balagopal, S.; Landro, T.; Sutija, D.P.

1997-01-01

Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate

Novel waste printed circuit board recycling process with molten salt.

Science.gov (United States)

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

Novel waste printed circuit board recycling process with molten salt

Science.gov (United States)

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

Fuel processing for molten-salt reactors

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1975-01-01

Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

Sulfomethylated lignite salt as a sacrifical agent in oil recovery processes

Energy Technology Data Exchange (ETDEWEB)

Kudchadker, M.V.; Weiss, W.J.

1978-02-07

A process is described for recovering petroleum from oil reservoirs by secondary recovery methods. The process involves injecting via an injection well into the reservoir an aqueous solution of sulfomethylated lignite salt as a sacrificial agent to inhibit the deposition of surfactant and polymer on the reservoir matrix. The process is conducted by first injecting the lignite salt into the formation through the injection well and following it with either a polymer or a surfactant solution, which also may contain the lignite salt. The polymer or surfactant would then be followed by a drive fluid, such as water, to push the chemicals and oil to the production well. (18 claims)

Cation exchange process for molten salt extraction residues

International Nuclear Information System (INIS)

Proctor, S.G.

1975-01-01

A new method, utilizing a cation exchange technique, has been developed for processing molten salt extraction (MSE) chloride salt residues. The developed ion exchange procedure has been used to separate americium and plutonium from gross quantities of magnesium, potassium, and sodium chloride that are present in the residues. The recovered plutonium and americium contained only 20 percent of the original amounts of magnesium, potassium, and sodium and were completely free of any detectable amounts of chloride impurity. (U.S.)

Liquid-liquid reductive extraction in molten fluoride salt/liquid aluminium as a core of process for the An/Ln group separation

International Nuclear Information System (INIS)

Conocar, O.

2007-06-01

This report concerns a pyrochemical process based on liquid-liquid extraction in a molten fluoride/liquid aluminium system as a core process for actinide (An)/lanthanide (Ln) group separation, studied at CEA. The basic and demonstrative experiments have established the feasibility of the An/Ln group separation in the molten fluoride/liquid aluminium system (U, Pu, Np, Am, Cm traces from Nd, Ce, Eu, Sm, Eu, La - An/Ln separation factors over 1000 - An recovery yield over 98 % in one batch). The main experimental efforts must now be targeted on the recovery of actinides from the Al matrix. A thermodynamic and bibliographical survey has been done. It shows that back-extraction in a molten chloride melt could be a promising technique for this purpose

Liquid-liquid reductive extraction in molten fluoride salt/liquid aluminium as a core of process for the An/Ln group separation

Energy Technology Data Exchange (ETDEWEB)

Conocar, O

2007-06-15

This report concerns a pyrochemical process based on liquid-liquid extraction in a molten fluoride/liquid aluminium system as a core process for actinide (An)/lanthanide (Ln) group separation, studied at CEA. The basic and demonstrative experiments have established the feasibility of the An/Ln group separation in the molten fluoride/liquid aluminium system (U, Pu, Np, Am, Cm traces from Nd, Ce, Eu, Sm, Eu, La - An/Ln separation factors over 1000 - An recovery yield over 98 % in one batch). The main experimental efforts must now be targeted on the recovery of actinides from the Al matrix. A thermodynamic and bibliographical survey has been done. It shows that back-extraction in a molten chloride melt could be a promising technique for this purpose.

Molten salt extractive distillation process for zirconium-hafnium separation

International Nuclear Information System (INIS)

McLaughlin, D.F.; Stoltz, R.A.

1989-01-01

This patent describes an improvement in a process for zirconium-hafnium separation. It utilizes an extractive distillation column with a mixture of zirconium and hafnium tetrachlorides introduced into a distillation column having a top and bottom with hafnium enriched overheads taken from the top of the column and a molten salt solvent circulated through the column to provide a liquid phase, and with molten salt solvent containing zirconium chloride being taken from the bottom of the distillation column. The improvements comprising: utilizing a molten salt solvent consisting principally of lithium chloride and at least one of sodium, potassium, magnesium and calcium chlorides; stripping of the zirconium chloride taken from the bottom of the distillation column by electrochemically reducing zirconium from the molten salt solvent; and utilizing a pressurized reflux condenser on the top of the column to add the hafnium chloride enriched overheads to the molten salt solvent previously stripped of zirconium chloride

Analysis of leachants from strontium chlorapatite glass ceramics

International Nuclear Information System (INIS)

Vijayalakshmi, S.; Ushalakshmi, K.; Annapoorani, S.; Sriram, S.; Uma Maheshwari, R.; Deivanayaki, R.; Sekar, J.K.; Sankaran, K.

2013-01-01

Strontium chlorapatite glass ceramics is being tried out as one of the candidate matrices for immobilizing pyrochemical salt waste produced in the nuclear industry. To find-out the suitability of such material for immobilising the waste, leaching of various constituents of the ceramics in water is required. Therefore, in Chemistry Group of IGCAR experiments are being carried out with simulated salt waste (chlorides of Li, Na, K, Cs, Ba, Nd and Ce) of pyrochemical reprocessing method for studying the utilisation of strontium chlorapatite glass ceramics towards the immobilization of radioactive waste. Leaching behaviour study requires the determination of alkali, alkaline earth and rare earth elements in the leachant solutions of the glass ceramic material. Apart from cations, leaching study of anions especially chloride is required as the chloride salts are used in pyrochemical experiments. Considering the good sensitivity of alkali elements in Flame-AES method, all the alkali elements were determined by flame-AES. Ba, Sr and rare earth elements in the leachant solutions were determined using ICP-OES. Chloride was determined using ISE and IC. Standardisation of instrumental techniques and the application of various techniques for the sample analysis will be discussed in the paper. (author)

Study of the actinide-lanthanide separation from nuclear waste by a new pyrochemical process; Etude de la separation actinides-lanthanides des dechets nucleaires par un procede pyrochimique nouveau

Energy Technology Data Exchange (ETDEWEB)

Lemort, F. [CEA Marcoule, Departement de Retraitement, des Dechets et du Demantelement, 30 - Bagnols-sur-Ceze (France)]|[Institut National Polytechnique, 38 - Grenoble (France)

1997-01-01

The theoretical extraction and separation of platinoids, actinides and lanthanides is allowed by thermodynamic using two adapted reducing agents: zinc and magnesium. Thereby, a pyrochemical method for the nuclear waste processing has been devised. The high temperature handling of the elements in fluoride forms and their processing by a reactive metallic phase required special precautions. The study of the behavior of matter in exploratory systems allowed the development of an experimental technology for the treatment and contacting of phases. The thermodynamical analysis of the experimental results shows the feasibility of the process. A model was developed to predict the distribution coefficients of zirconium, uranium and lanthanum as a function of the system composition. An estimation method was proposed in order to evaluate the distribution coefficients in diluted solution of all the actinides and lanthanides existing in the fission products between LiF CaF{sub 2} and Zn-Mg at 720 deg C. Coupled with the experimental results, the estimates results may be extrapolated to concentrated solutions allowing predictions of the separation of all actinides and lanthanides. The rapidity of element transfer is induced by a thermal effect caused by the high exothermicity of the reduction by magnesium. The kinetic coefficients have been linked with the reduction enthalpy of each element. Moreover, the kinetics seem limited by chemical reaction and not by mass transfer. (author) 66 refs.

Molten salt destruction process for mixed wastes

International Nuclear Information System (INIS)

Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

1993-04-01

We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

Pyro-chemistry within the FP7 ACSEPT Project-Program and Objective

Energy Technology Data Exchange (ETDEWEB)

Caravaca, Concha [CIEMAT/Nuclear Fission Division/URAA, Avda. Complutense, 22.Madrid 28040 (Spain); Bourg, Stephane [CEA/DEN/MAR/DRCP, CEA Marcoule. BP17171, 30207 Bagnols sur Ceze Cedex (France)

2008-07-01

Actinide recycling by partitioning and transmutation is considered as one of the most promising strategies to reduce the inventory of radioactive waste, thus contributing to make nuclear energy sustainable. To make advances beyond the current state of the art in pyrochemical separations processes, the Domain 2 (DM2) of ACSEPT has been built on considering a process approach based on system studied. Four work packages that represent the main steps of a process block diagram have been identified: head-end steps, core process development, and salt treatment for recycling and waste conditioning. The results obtained in this domain will be integrated in DM 3 (Process) in order to orientate the R and D studies of DM2 and to propose and validate flowsheets at the end of the project. The state of the art on pyrochemical separation within the European Community and the working program of ACSEPT in pyrometallurgy are presented in this work. (authors)

Novel waste printed circuit board recycling process with molten salt

OpenAIRE

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450?470??C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, a...

Electrodialysis-based separation process for salt recovery and recycling from waste water

Science.gov (United States)

Tsai, Shih-Perng

1997-01-01

A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

Leach resistance properties and release processes for salt-occluded zeolite A

International Nuclear Information System (INIS)

Lewis, M.A.; Fischer, D.F.; Laidler, J.J.

1992-01-01

The pyrometallurgical processing of spent fuel from the Integral Fast Reactor (IFR) results in a waste of LiCl-KCl-NaCl salt containing approximately 10 wt% fission products, primarily CsCl and SrCl 2 . For disposal, this waste must be immobilized in a form that it is leach resistant. A salt-occluded zeolite has been identified as a potential waste form for the salt. Its leach resistance properties were investigated using powdered samples. The results were that strontium was not released and cesium had a low release, 0.056 g/m 2 for the 56 day leach test. The initial release (within 7 days) of alkali metal cations was rapid and subsequent releases were much smaller. The releases of aluminum and silicon were 0.036 and 0.028 g/m 2 , respectively, and were constant. Neither alkali metal cation hydrolysis nor exchange between cations in the leachate and those in the zeolite was significant. Only sodium release followed t 0.5 kinetics. Selected dissolution of the occluded salt was the primary release process. These results confirm that salt-occluded zeolite has promise as the waste form for IFR pyroprocess salt

Fluoride partitioning R and D programme for molten salt transmutation reactor systems in the Czech Republic

International Nuclear Information System (INIS)

Uhlir, J.; Priman, V.; Vanicek, J.

2001-01-01

The transmutation of spent nuclear fuel is considered a prospective alternative conception to the current conception based on the non-reprocessed spent fuel disposal into underground repository. The Czech research and development programme in the field of partitioning and transmutation is founded on the Molten Salt Transmutation Reactor system concept with fluoride salts based liquid fuel, the fuel cycle of which is grounded on pyrochemical / pyrometallurgical fluoride partitioning of spent fuel. The main research activities in the field of fluoride partitioning are oriented mainly towards technological research of Fluoride Volatility Method and laboratory research on electro-separation methods from fluoride melts media. The Czech national conception in the area of P and T research issues from the national power industry programme and from the Czech Power Company intentions of the extensive utilization of nuclear power in our country. The experimental R and D work is concentrated mainly in the Nuclear Research Institute Rez plc that plays a role of main nuclear research workplace for the Czech Power Company. (author)

A novel bread making process using salt-stressed Baker's yeast.

Science.gov (United States)

Yeh, Lien-Te; Charles, Albert Linton; Ho, Chi-Tang; Huang, Tzou-Chi

2009-01-01

By adjusting the mixing order of ingredients in traditional formula, an innovative bread making process was developed. The effect of salt-stressed Baker's yeast on bread dough of different sugar levels was investigated. Baker's yeast was stressed in 7% salt solution then mixed into dough, which was then evaluated for fermentation time, dough fermentation producing gas, dough expansion, bread specific volumes, and sensory and physical properties. The results of this study indicated that salt-stressed Baker's yeast shortened fermentation time in 16% and 24% sugar dough. Forty minutes of salt stress produced significant amount of gas and increased bread specific volumes. The bread was softer and significantly improved sensory properties for aroma, taste, and overall acceptability were obtained.

Molten salt corrosion behavior of structural materials in LiCl-KCl-UCl3 by thermogravimetric study

Science.gov (United States)

Rao, Ch Jagadeeswara; Ningshen, S.; Mallika, C.; Mudali, U. Kamachi

2018-04-01

The corrosion resistance of structural materials has been recognized as a key issue in the various unit operations such as salt purification, electrorefining, cathode processing and injection casting in the pyrochemical reprocessing of spent metallic nuclear fuels. In the present work, the corrosion behavior of the candidate materials of stainless steel (SS) 410, 2.25Cr-1Mo and 9Cr-1Mo steels was investigated in molten LiCl-KCl-UCl3 salt by thermogravimetric analysis under inert and reactive atmospheres at 500 and 600 °C, for 6 h duration. Insignificant weight gain (less than 1 mg/cm2) in the inert atmosphere and marginal weight gain (maximum 5 mg/cm2) in the reactive atmosphere were observed at both the temperatures. Chromium depletion rates and formation of Cr-rich corrosion products increased with increasing temperature of exposure in both inert and reactive atmospheres as evidenced by SEM and EDS analysis. The corrosion attack by LiCl-KCl-UCl3 molten salt, under reactive atmosphere for 6 h duration was more in the case of SS410 than 9Cr-1Mo steel followed by 2.25Cr-1Mo steel at 500 °C and the corrosion attack at 600 °C followed the order: 9Cr-1Mo steel >2.25Cr-1Mo steel > SS410. Outward diffusion of the minor alloying element, Mo was observed in 9Cr-1Mo and 2.25Cr-1Mo steels at both temperatures under reactive atmosphere. Laser Raman spectral analysis of the molten salt corrosion tested alloys under a reactive atmosphere at 500 and 600 °C for 6 h revealed the formation of unprotected Fe3O4 and α-as well as γ-Fe2O3. The results of the present study facilitate the selection of structural materials for applications in the corrosive molten salt environment at high temperatures.

Study of electrochemical processes for separation of the actinides and lanthanides in molten fluoride media

International Nuclear Information System (INIS)

Zvejskova, R.; Chuchvalcova Bimova, K.; Lisy, F.; Soucek, P.

2005-01-01

The technology of the Molten Salt Reactors (MSR) is developed for two possible applications: For one thing as the Molten Salt Transmutation Reactor (MSTR) incinerating plutonium and minor actinides within reprocessing of spent fuel from PWR or FBR and for another thing as electricity generating MSR working under thorium uranium fuel cycle. Electrochemical separation processes are one of promising pyrochemical techniques that should enable the on-line reprocessing of circulating fuel salt in MSR (fuel cycle back-end). The former application represents the Czech P and T concept, in which framework the electrolytic separation can be applied both in the front-end and back-end of the MSTR fuel cycle. Within the front-end electro separation should follow the Fluoride Volatility Method (FVM), which should separate 95 % of uranium from the spent fuel in the form of volatile uranium hexafluoride. The residual uranium and fission products (FP) are supposed to be separated among others also by electrochemical methods. The presented work comprises the results reached within development of electrochemical separation of the actinides and fission products from each other by electrolytic deposition method on solid cathode in molten fluoride media, that represent he carrier salts of MSR technology. The knowledge of electrochemical properties (red-ox potentials, mainly of deposition potentials) is necessary for determination of separation possibilities of individual components by electrolysis. (authors)

Actinide-Lanthanide separation by an electrolytic method in molten salt media: feasibility assessment of a renewed liquid cathode

International Nuclear Information System (INIS)

Huguet, A.

2009-12-01

This study is part of a research program concerning the assessment of pyrochemical methods for the nuclear waste processing. The An-Ln partitioning could be achieved by an electrolytic selective extraction in molten salt media. It has been decided to focus on liquid reactive cathode which better suits to a group actinides co-recycling. The aim of the study is to propose, define and initiate the development of an electrolytic pyro-process dedicated to the quantitative and selective recovery of the actinides. Quantitativeness is related to technology, whereas selectivity is governed by chemistry. The first step consisted in selecting the adequate operating conditions, which enables a sufficient An-Ln separation. The first step consisted, by means of thermodynamic calculi and electrochemical investigations, in selecting a promising combination between molten electrolyte and cathodic material, regarding the process constraints. To improve the recovery yield, it is necessary to develop a disruptive technology: here comes the concept of a dynamic electrodeposition carried out onto liquid metallic drops. The next step consisted in designing and manufacturing a lab-scale device which enables dropping flow studies. Since interfacial phenomena are of primary meaning in such a concept, it has been decided to focus on high temperature liquid-liquid interfacial measurements. (author)

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO2-Al2O3-P2O5(SAP) Inorganic Composite: Part 2. The Effect of SAP Composition on Stabilization/Solidification

International Nuclear Information System (INIS)

Ahn, Soo Na; Park, Hwan Seo; Cho, In Hak; Kim, In Tae; Cho, Yong Zun

2012-01-01

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe 2 O 3 into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP(Fe=0.1). The experimental results indicated that the addition of Fe 2 O 3 increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B 2 O 3 into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 - 4 g of wasteform for final disposal. The final volume would be about 3 - 4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

Testing of pyrochemical centrifugal contactors

International Nuclear Information System (INIS)

Chow, L.S.; Carls, E.L.; Basco, J.K.; Johnson, T.R.

1996-01-01

A centrifugal contactor that performs oxidation and reduction exchange reactions between molten metals and salts at 500 degrees Centigrade has been tested successfully at Argonne National Laboratory (ANL). The design is based on contactors for aqueous- organic systems operation near room temperature. In tests to demonstrate the performance of the pyrocontactor, cadmium and LICl-KCl eutectic salt were the immiscible solvent phases, and rare earths were the distributing solutes. The tests showed that the pyrocontactor mixed and separated the phases well, with stage efficiencies approaching 99% at rotor speeds near 2700 rpm. The contactor ran smoothly and reliably over the entire range of speeds that was tested

ALTERNATIVE METHODS OF TECHNOLOGICAL PROCESSING TO REDUCE SALT IN MEAT PRODUCTS

Directory of Open Access Journals (Sweden)

E. K. Tunieva

2017-01-01

Full Text Available The world trends in table salt reduction in meat products contemplate the use of different methods for preservation of taste and consistency in finished products as well as shelf life prolongation. There are several approaches to a sodium chloride reduction in meat products. The paper presents a review of the foreign studies that give evidence of the possibility to maintain quality of traditional meat products produced with the reduced salt content. The studies in the field of salty taste perception established that a decrease in a salt crystal size to 20 µm enabled reducing an amount of added table salt due to an increase in the salty taste intensity in food products. Investigation of the compatibility of different taste directions is also interesting as one of the approaches to a sodium chloride reduction in food products. The use of water-in-oil-in-water (w/o/w double emulsions allows controlling a release of encapsulated ingredients (salt, which enables enhancement of salty taste. The other alternative method of technological processing of meat raw material for reducing salt in meat products is the use of high pressure processing. This method has several advantages and allows not only an increase in the salty taste intensity, but also formation of a stable emulsion, an increase in water binding capacity of minced meat and extension of shelf-life.

Waste management in IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.; Battles, J.E.

1991-01-01

The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

Evaluation of alkali bromide salts for potential pyrochemical applications

International Nuclear Information System (INIS)

Tripathy, P.K.; Gutknecht, T.Y.; Herrmann, S.D.; Fredrickson, G.L.; Lister, T.E.

2013-01-01

Transient techniques were employed to study the electrochemical behavior, reduction mechanism and transport properties of REBr 3 (RE - La, Nd and Gd) in pure LiBr, LiBr-KBr (eutectic) and LiBr-KBr-CsBr (eutectic) melts. Gd(III) showed a reversible single step soluble-insoluble exchange phenomenon in LiBr melt at 973 K. Although La (III), Nd(III) and Gd(III) ions showed reversible behavior in eutectic LiBr-KBr melts, these ions showed a combination of temperature dependent reversible and pseudo-reversible behavior. While both La(III) and Gd(III) showed one step reduction, the reduction of Nd(III) was observed to be a two step process. La metal could be electrodeposited from the ternary electrolyte at a temperature of 673 K. Various electrochemical measurements suggest that both binary and ternary bromide melts can potentially be used to electro-deposit high purity RE metals at comparatively lower operating temperatures. (authors)

Evaluation of Alkali Bromide Salts for Potential Pyrochemical Applications

Energy Technology Data Exchange (ETDEWEB)

Prabhat K. Tripathy; Steven D. Herrmann; Guy L. Fredrickson; Tedd E. Lister; Toni Y. Gutknecht

2013-10-01

Transient techniques were employed to study the electrochemical behavior, reduction mechanism and transport properties of REBr3 (RE - La, Nd and Gd) in pure LiBr, LiBr-KBr (eutectic) and LiBr-KBr-CsBr (eutectic) melts. Gd(III) showed a reversible single step soluble-insoluble exchange phenomenon in LiBr melt at 973K. Although La (III), Nd(III) and Gd(III) ions showed reversible behavior in eutectic LiBr-KBr melts, these ions showed a combination of temperature dependent reversible and pseudo-reversible behavior. While both La(III) and Gd(III) showed one step reduction, the reduction of Nd(III) was observed to be a two step process. La metal could be electrodeposited from the ternary electrolyte at a temperature of 673K. Various electrochemical measurements suggest that both binary and ternary bromide melts can potentially be used to electrodeposit high purity RE metals at comparatively lower operating temperatures.

Comparison of phosgene, chlorine, and hydrogen chloride as reagents for converting molten CaO.CaCl2 to CaCl2

International Nuclear Information System (INIS)

Fife, K.W.

1985-01-01

One method at Los Alamos for preparing impure plutonium metal from the impure oxide is by batch reduction with calcium metal at 850 0 C in a CaCl 2 solvent. The solvent salt from this reduction is currently discarded as low-level radioactivity waste only because it is saturated with the CaO byproduct. We have demonstrated a pyrochemical technique for converting the CaO to CaCl 2 thereby incorporating solvent recycling into the batch reduction process. We will discuss the effectiveness of HCl, Cl 2 , and COCl 2 as chlorinating agents and recycling actual spent process solvent salts. 6 refs., 8 figs

Influence of salt content and processing time on sensory characteristics of cooked "lacón".

Science.gov (United States)

Purriños, Laura; Bermúdez, Roberto; Temperán, Sara; Franco, Daniel; Carballo, Javier; Lorenzo, José M

2011-04-01

The influence of salt content and processing time on the sensory properties of cooked "lacón" were determined. "Lacón" is a traditional dry-cured and ripened meat product made in the north-west of Spain from the fore leg of the pig, following a similar process to that of dry-cured ham. Six batches of "lacón" were salted with different amounts of salt (LS (3 days of salting), MS (4 days of salting) and HS (5 days of salting)) and ripened during two times (56 and 84 days of dry-ripening). Cured odour in all batches studied, red colour and rancid odour in MS and HS batches, flavour intensity in MS batch and fat yellowness, rancid flavour and hardness in the HS batch were significantly different with respect to the time of processing. Appearance, odour, flavour and texture were not significantly affected by the salt content (P>0.05). However, the saltiness score showed significant differences with respect to the salt levels in all studied batches (56 and 84 days of process). The principal component analysis showed that physicochemical traits were the most important ones concerning the quality of dry-cured "lacón" and offered a good separation of the mean samples according to the dry ripening days and salt level. © 2010 The American Meat Science Association. Published by Elsevier Ltd. All rights reserved.

Process for the recovery of alkali metal salts from aqueous solutions thereof

International Nuclear Information System (INIS)

Vitner, J.

1984-01-01

In an integrated process for the recovery of alkakli metal phenates and carboxylates from aqueous solutions thereof, the aqueous solution is spray dried and the drying gas stream is contacted with an aqueous alkali metal salt solution which dissolves the particles of the alkali metal salt that were entrained in the drying gas stream. The salt-free inert gas stream is then dried, heated, and returned to the spray dryer

Targets and timelines for reducing salt in processed food in the Americas.

Science.gov (United States)

Campbell, Norm; Legowski, Barbara; Legetic, Branka; Ferrante, Daniel; Nilson, Eduardo; Campbell, Christine; L'Abbé, Mary

2014-09-01

Reducing dietary salt is one of the most effective interventions to lessen the burden of premature death and disability. In high-income countries and those in nutrition transition, processed foods are a significant if not the main source of dietary salt. Reformulating these products to reduce their salt content is recommended as a best buy to prevent chronic diseases across populations. In the Americas, there are targets and timelines for reduced salt content of processed foods in 8 countries--Argentina, Brazil, Canada, Chile, Ecuador, Mexico, and the National Salt Reduction Initiative in the United States and Paraguay. While there are common elements across the countries, there are notable differences in their approaches: 4 countries have exclusively voluntary targets, 2 countries have combined voluntary and regulated components, and 1 country has only regulations. The countries have set different types of targets and in some cases combined them: averages, sales-weighted averages, upper limits, and percentage reductions. The foods to which the targets apply vary from single categories to comprehensive categories accounting for all processed products. The most accessible and transparent targets are upper limits per food category. Most likely to have a substantive and sustained impact on salt intake across whole populations is the combination of sales-weighted averages and upper limits. To assist all countries with policies to improve the overall nutritional value of processed foods, the authors call for food companies to supply food composition data and product sales volume data to transparent and open-access platforms and for global companies to supply the products that meet the strictest targets to all markets. Countries participating in common markets at the subregional level can consider harmonizing targets, nutrition labels, and warning labels. ©2014 Wiley Periodicals, Inc.

Recycle of LWR actinides to an IFR

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)

Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

International Nuclear Information System (INIS)

Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

2002-01-01

The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}(SAP) Inorganic Composite: Part 2. The Effect of SAP Composition on Stabilization/Solidification

Energy Technology Data Exchange (ETDEWEB)

Ahn, Soo Na; Park, Hwan Seo; Cho, In Hak; Kim, In Tae; Cho, Yong Zun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-03-15

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe{sub 2}O{sub 3} into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP(Fe=0.1). The experimental results indicated that the addition of Fe{sub 2}O{sub 3} increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B{sub 2}O{sub 3} into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 - 4 g of wasteform for final disposal. The final volume would be about 3 - 4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

Use of alternative curing salts for processing salamis

Directory of Open Access Journals (Sweden)

Dong-Gyun Yim

2018-01-01

Full Text Available Objective This study was performed to determine effects of different curing salts on the quality of salamis and to assess feasibility of using NaCl-alternative salts. Methods Various types of curing salts (KCl or MgCl2 as well as NaCl (sun-dried or refined were incorporated for processing of salamis. The proximate composition, fatty acids, nucleotide-related compounds, and free amino acids of the salamis were analyzed during 40 days of ripening. Results The substitution of NaCl by KCl caused higher fat and ash content, but lower moisture content of the salami after 20 days of ripening (p<0.05. Compared with the sun-dried NaCl, use of KCl in salami also led to greater inosine 5′-monophosphate whereas refined NaCl had more inosine (p<0.05. KCl-added salami also had a higher C12:0, C17:1, and C20:0 than other types of salami (p<0.05. MgCl2-added salami had higher content of free amino acids compared to the other salamis (p<0.05. Conclusion Alternative curing salts such as KCl and MgCl2 could substitute NaCl in consideration of quality factor of a fermented meat product. Especially replacement of NaCl with KCl will be a suitable strategy for developing relatively low sodium salami products without compromising product quality.

Engineering development studies for molten-salt breeder reactor processing No. 21

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1976-03-01

The status of the following programs is reported: (1) continuous fluorinator development: autoresistance heating test AHT-4; (2) development of the metal transfer process; (3) salt-metal contactor development: experiments with a mechanically agitated, nondispersing contactor using water and mercury and in the salt-bismuth flowthrough facility; and (4) fuel reconstitution development: installation of equipment for a fuel reconstitution engineering experiment

Recycling of Uranium from Uranium-Aluminium alloys by Chlorination with HCl(g)

OpenAIRE

MEIER ROLAND; SOUCEK Pavel; MALMBECK Rikard; FANGHAENEL Thomas

2012-01-01

Much attention has been paid to the handling of metallic nuclear fuel, which can offer a safe and more secure fuel cycle. To ensure this, it is essential to gain a thorough understanding of the related fundamental processes to scientifically assess the safety aspects. To this aim, a pyrochemical process for the recycling of actinides is being investigated in ITU. It is consisting of anodic dissolution of all actinides to a molten salt bath and electrochemical reduction on solid aluminium cath...

Magnetic separation as a plutonium residue enrichment process

International Nuclear Information System (INIS)

Avens, L.R.; Gallegos, U.F.; McFarlan, J.T.

1990-01-01

Several plutonium contaminated residues have been subjected to Open Gradient Magnetic Separation (OGMS) on an experimental scale. OGMS experiments on graphite and bomb reduction residues resulted in a plutonium rich fraction and a plutonium lean fraction. Values for the bulk quantity rejected to the lean fraction varied between about 20% to 85% of the feed bulk. The plutonium content of the lean fraction can be reduced from about 2% in the feed to the 0.1% to 0.5% range dependent on the portion of the feed rejected to this lean fraction. These values are low enough in plutonium to meet economic discard limits and be considered for direct discard. Magnetic separation of pyrochemical salts gave less favorable results. While a fraction very rich in plutonium could be obtained, the lean fraction plutonium content was too high for direct discard. This may still have chemical processing applications. OGMS experiments at low magnetic field strength on incinerator ash did give two fractions but the plutonium content of each fraction was essentially identical. Thus, no chemical processing advantage was identified for magnetic separation of this residue. 6 refs., 1 fig., 9 tabs

The electrochemical reduction processes of solid compounds in high temperature molten salts.

Science.gov (United States)

Xiao, Wei; Wang, Dihua

2014-05-21

Solid electrode processes fall in the central focus of electrochemistry due to their broad-based applications in electrochemical energy storage/conversion devices, sensors and electrochemical preparation. The electrolytic production of metals, alloys, semiconductors and oxides via the electrochemical reduction of solid compounds (especially solid oxides) in high temperature molten salts has been well demonstrated to be an effective and environmentally friendly process for refractory metal extraction, functional materials preparation as well as spent fuel reprocessing. The (electro)chemical reduction of solid compounds under cathodic polarizations generally accompanies a variety of changes at the cathode/melt electrochemical interface which result in diverse electrolytic products with different compositions, morphologies and microstructures. This report summarizes various (electro)chemical reactions taking place at the compound cathode/melt interface during the electrochemical reduction of solid compounds in molten salts, which mainly include: (1) the direct electro-deoxidation of solid oxides; (2) the deposition of the active metal together with the electrochemical reduction of solid oxides; (3) the electro-inclusion of cations from molten salts; (4) the dissolution-electrodeposition process, and (5) the electron hopping process and carbon deposition with the utilization of carbon-based anodes. The implications of the forenamed cathodic reactions on the energy efficiency, chemical compositions and microstructures of the electrolytic products are also discussed. We hope that a comprehensive understanding of the cathodic processes during the electrochemical reduction of solid compounds in molten salts could form a basis for developing a clean, energy efficient and affordable production process for advanced/engineering materials.

Process technology for the molten-salt reactor 233U--Th cycle

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1975-01-01

After a brief description of the design features of the molten-salt breeder reactor, fuel processing for removal of 233 Pa and fission products is examined. Some recent developments in processing technology are discussed

Actinide removal from molten salts by chemical oxidation and salt distillation

Energy Technology Data Exchange (ETDEWEB)

McNeese, J.A.; Garcia, E.; Dole, V.R. [Los Alamos National Laboratory, NM (United States)] [and others

1995-10-01

Actinide removal from molten salts can be accomplished by a two step process where the actinide is first oxidized to the oxide using a chemical oxidant such as calcium carbonate or sodium carbonate. After the actinide is precipitated as an oxide the molten salt is distilled away from the actinide oxides leaving a oxide powder heel and an actinide free distilled salt that can be recycled back into the processing stream. This paper discusses the chemistry of the oxidation process and the physical conditions required to accomplish a salt distillation. Possible application of an analogous process sequence for a proposed accelerator driven transmutation molten salt process is also discussed.

Actinide removal from molten salts by chemical oxidation and salt distillation

International Nuclear Information System (INIS)

McNeese, James A.; Garcia, Eduardo; Dole, Vonda R.; Griego, Walter J.

1995-01-01

Actinide removal from molten salts can be accomplished by a two step process where the actinide is first oxidized to the oxide using a chemical oxidant such as calcium carbonate or sodium carbonate. After the actinide is precipitated as an oxide the molten salt is distilled away from the actinide oxides leaving a oxide powder heel and an actinide free distilled salt that can be recycled back into the processing stream. This paper discusses the chemistry of the oxidation process and the physical conditions required to accomplish a salt distillation. Possible application of an analogous process sequence for a proposed accelerator driven transmutation molten salt process is also discussed

Concept of the plant for the BN-800 fast reactor fuel recycling with application of pyro-process and vibro-packing technology

International Nuclear Information System (INIS)

Bychkov, A.V.; Skiba, O.V.; Mayorshin, A.A.; Demidova, L.S.; Kormilitzyna, L.A.; Ishunin, V.S.

2000-01-01

The conception of Plant was developed for MOX-fuel recycle at two BN-800 type fast reactors by pyrochemical reprocessing of irradiated nuclear fuel (INF) and production of vibro-pac fuel pins and SA. INF production process and stages of pyrochemical reprocessing were analyzed. Starting materials were chosen. Characteristics of irradiated SA and requirements for finished products were defined. Volumes of production were estimated. Procedure of waste management was defined. The following description was made: (1) general flow sheet of fuel recycling and partial schemes of single reprocessing; (2) composition of production process equipment; (3) arrangement of production process equipment; (4) lay out of Plant building and engineering communications. Principle economical assessments were made for production under design. (authors)

Cathodic processes in high-temperature molten salts for the development of new materials processing methods

International Nuclear Information System (INIS)

Schwandt, Carsten

2017-01-01

Molten salts play an important role in the processing of a range of commodity materials. This includes the large-scale production of iron, aluminium, magnesium and alkali metals as well as the refining of nuclear fuel materials. This presentation focuses on two more recent concepts in which the cathodic reactions in molten salt electrolytic cells are used to prepare high-value-added materials. Both were developed and advanced at the Department of Materials Science and Metallurgy at the University of Cambridge and are still actively being pursued. One concept is now generally known as the FFC-Cambridge process. The presentation will highlight the optimisation of the process towards high selectivities for tubes or particles depict a modification of the method to synthesize tin-filled carbon nanomaterial, and illustrate the implementation of a novel type of process control to enable the preparation of gramme quantities of material within a few hours with simple laboratory equipment. Also discussed will be the testing of these materials in lithium ion batteries

Pyrochemical recovery of easily reducible species from spent nuclear fuel

International Nuclear Information System (INIS)

Jouault, C.

2000-01-01

The purpose of the reprocessing of spent fuel is to separate noble metals and other easily reducible species, actinides and lanthanides. A thermodynamic and bibliographical study allowed us to elaborate a process which realises these separations in several steps. The experimental validation of the steps concerning the extraction of noble metals and easily reducible species required to imagine an apparatus which is conformed to the study of the two steps in question: the reduction by a gas of fission product oxides and the extraction of the metallic particles, obtained by reduction, by digestion in a liquid metal. Experiments on digestion, carried on molybdenum and ruthenium particles, allowed us to conclude that the transfer of metallic particles from a molten salt into a liquid metal is ruled by phenomena of complex wettability between the metallic particle, the molten salt, the liquid metal and the gas. The transfer from the salt to the metal is a chain of two steps: emersion of the particles from the salt to go into the gas, and then transfer from the gas into the metal. Kinetics are limited by the transfer through the metal surface. Kinetics study withdrew the experimental parameters and the metals properties which influence the digestion rate. A model on the transfer into a liquid metal of a particle trapped at the fluid/metal interface ratified the experimental conclusions and informed on the stirring influence. All the results allow us to think that the extraction of noble metals and easily reducible species are feasible in this way. (author) [fr

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

Energy Technology Data Exchange (ETDEWEB)

Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-08-08

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

Treatment of radioactive waste salt by using synthetic silica-based phosphate composite for de-chlorination and solidification

Science.gov (United States)

Cho, In-Hak; Park, Hwan-Seo; Lee, Ki-Rak; Choi, Jung-Hun; Kim, In-Tae; Hur, Jin Mok; Lee, Young-Seak

2017-09-01

In the radioactive waste management, waste salts as metal chloride generated from a pyrochemical process to recover uranium and transuranic elements are one of problematic wastes due to their intrinsic properties such as high volatility and low compatibility with conventional glasses. This study reports a method to stabilize and solidify LiCl waste via de-chlorination using a synthetic composite, U-SAP (SiO2-Al2O3-B2O3-Fe2O3-P2O5) prepared by a sol-gel process. The composite was reacted with alkali metal elements to produce some metal aluminosilicates, aluminophosphates or orthophosphate as a crystalline or amorphous compound. Different from the original SAP (SiO2-Al2O3-P2O5), the reaction product of U-SAP could be successfully fabricated as a monolithic wasteform without a glassy binder at a proper reaction/consolidation condition. From the results of the FE-SEM, FT-IR and MAS-NMR analysis, it could be inferred that the Si-rich phase and P-rich phase as a glassy grains would be distributed in tens of nm scale, where alkali metal elements would be chemically interacted with Si-rich or P-rich region in the virgin U-SAP composite and its products was vitrified into a silicate or phosphate glass after a heat-treatment at 1150 °C. The PCT-A (Product Consistency Test, ASTM-1208) revealed that the mass loss of Cs and Sr in the U-SAP wasteform had a range of 10-3∼10-1 g/m2 and the leach-resistance of the U-SAP wasteform was comparable to other conventional wasteforms. From the U-SAP method, LiCl waste salt was effectively stabilized and solidified with high waste loading and good leach-resistance.

Modeling Coupled THMC Processes and Brine Migration in Salt at High Temperatures

Energy Technology Data Exchange (ETDEWEB)

Rutqvist, Jonny; Blanco Martin, Laura; Mukhopadhyay, Sumit; Houseworth, Jim; Birkholzer, Jens

2014-08-14

In this report, we present FY2014 progress by Lawrence Berkeley National Laboratory (LBNL) related to modeling of coupled thermal-hydrological-mechanical-chemical (THMC) processes in salt and their effect on brine migration at high temperatures. LBNL’s work on the modeling of coupled THMC processes in salt was initiated in FY2012, focusing on exploring and demonstrating the capabilities of an existing LBNL modeling tool (TOUGH-FLAC) for simulating temperature-driven coupled flow and geomechanical processes in salt. This work includes development related to, and implementation of, essential capabilities, as well as testing the model against relevant information and published experimental data related to the fate and transport of water. we provide more details on the FY2014 work, first presenting updated tools and improvements made to the TOUGH-FLAC simulator, and the use of this updated tool in a new model simulation of long-term THM behavior within a generic repository in a salt formation. This is followed by the description of current benchmarking and validations efforts, including the TSDE experiment. We then present the current status in the development of constitutive relationships and the dual-continuum model for brine migration. We conclude with an outlook for FY2015, which will be much focused on model validation against field experiments and on the use of the model for the design studies related to a proposed heater experiment.

A daily salt balance model for stream salinity generation processes following partial clearing from forest to pasture

Directory of Open Access Journals (Sweden)

M. A. Bari

2006-01-01

Full Text Available We developed a coupled salt and water balance model to represent the stream salinity generation process following land use changes. The conceptual model consists of three main components with five stores: (i Dry, Wet and Subsurface Stores, (ii a saturated Groundwater Store and (iii a transient Stream zone Store. The Dry and Wet Stores represent the salt and water movement in the unsaturated zone and also the near-stream dynamic saturated areas, responsible for the generation of salt flux associated with surface runoff and interflow. The unsaturated Subsurface Store represents the salt bulge and the salt fluxes. The Groundwater Store comes into play when the groundwater level is at or above the stream invert and quantifies the salt fluxes to the Stream zone Store. In the stream zone module, we consider a 'free mixing' between the salt brought about by surface runoff, interflow and groundwater flow. Salt accumulation on the surface due to evaporation and its flushing by initial winter flow is also incorporated in the Stream zone Store. The salt balance model was calibrated sequentially following successful application of the water balance model. Initial salt stores were estimated from measured salt profile data. We incorporated two lumped parameters to represent the complex chemical processes like diffusion-dilution-dispersion and salt fluxes due to preferential flow. The model has performed very well in simulating stream salinity generation processes observed at Ernies and Lemon experimental catchments in south west of Western Australia. The simulated and observed stream salinity and salt loads compare very well throughout the study period with NSE of 0.7 and 0.4 for Ernies and Lemon catchment respectively. The model slightly over predicted annual stream salt load by 6.2% and 6.8%.

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

International Nuclear Information System (INIS)

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

Vacuum evaporation of KCl-NaCl salts. Part 2: Vaporization-rate model and experimental results

International Nuclear Information System (INIS)

Wang, L.L.; Wallace, T.C. Sr.; Hampel, F.G.; Steele, J.H.

1996-01-01

Separation of chloride salts from the actinide residue by vacuum evaporation is a promising method of treating wastes from the pyrochemical plutonium processes. A model based on the Hertz-Langmuir relation is used to describe how evaporation rates of the binary KCl-NaCl system change with time. The effective evaporation coefficient (α), which is a ratio of the actual evaporation rate to the theoretical maximum, was obtained for the KCl-NaCl system using this model. In the temperature range of 640 C to 760 C, the effective evaporation coefficient ranges from ∼0.4 to 0.1 for evaporation experiments conducted at 0.13 Pa. At temperatures below the melting point, the lower evaporation coefficients are suggested to result from the more complex path that a molecule needs to follow before escaping to the gas phase. At the higher liquid temperatures, the decreasing evaporation coefficients result from a combination of the increasing vapor-flow resistances and the heat-transfer effects at the evaporation surface and the condensate layer. The microanalysis of the condensate verified that composition of the condensate changes with time, consistent with the model calculation. The microstructural examination revealed that the vaporate may have condensed as a single solution phase, which upon cooling forms fine lamellar structures of the equilibrium KCl and NaCl phases. In conclusion, the optimum design of the evaporation process and equipment must take the mass and heat transfer factors and equipment materials issues into consideration

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

International Nuclear Information System (INIS)

Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

2016-01-01

This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na₂O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

2011-04-01

The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

International Nuclear Information System (INIS)

Sabharwall, Piyush; Kim, Eung Soo; McKellar, Michael; Anderson, Nolan

2011-01-01

The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-13

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-18

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Use of sodium salt electrolysis in the process of continuous ...

Indian Academy of Sciences (India)

Use of sodium salt electrolysis in the process of continuous modification of eutectic EN ... the plastic groundmass of the solid solution α (Al), have an effect on their ..... Onyia C, Okorie B, Neife S and Obayi C 2013 World J. Eng. Technol. 1 9. 35.

Separation and Fixation of Toxic Components in Salt Brines Using a Water-Based Process

International Nuclear Information System (INIS)

Franks, Carrie J.; Quach, Anh P.; Birnie, Dunbar P.; Ela, Wendell P.; Saez, Avelino E.; Zelinski, Brian J.; Smith, Harry D.; Smith, Gary Lynn L.

2004-01-01

Efforts to implement new water quality standards, increase water reuse and reclamation, and minimize the cost of waste storage motivate the development of new processes for stabilizing waste water residuals that minimize waste volume, water content and the long-term environmental risk from related by products. This work explores the use of an aqueous-based emulsion process to create an epoxy/rubber matrix for separating and encapsulating waste components from salt laden, arsenic contaminated, amorphous iron hydrate sludges. Such sludges are generated from conventional water purification precipitation/adsorption processes, used to convert aqueous brine streams to semi-solid waste streams, such as ion exchange/membrane separation, and from other precipitative heavy metal removal operations. In this study, epoxy and polystyrene butadiene (PSB) rubber emulsions are mixed together and then combined with a surrogate sludge. The surrogate sludge consists of amorphous iron hydrate with 1 part arsenic fixed to the surface of the hydrate per 10 parts iron mixed with sodium nitrate and chloride salts and water. The resulting emulsion is cured and dried at 80 C to remove water. Microstructure characterization by electron microscopy confirms that the epoxy/PSB matrix surrounds and encapsulates the arsenic laden amorphous iron hydrate phase while allowing the salt to migrate to internal and external surfaces of the sample. Salt extraction studies indicate that the porous nature of the resulting matrix promotes the separation and removal of as much as 90% of the original salt content in only one hours time. Long term leaching studies based on the use of the infinite slab diffusion model reveal no evidence of iron migration or, by inference, arsenic migration, and demonstrate that the diffusion coefficients of the unextracted salt yield leachability indices within regulations for non-hazardous landfill disposal. Because salt is the most mobile species, it is inferred that arsenic

Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

International Nuclear Information System (INIS)

Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

2008-01-01

In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

Salt-assisted and salt-suppressed sol-gel transitions of methylcellulose in water.

Science.gov (United States)

Xu, Y; Wang, C; Tam, K C; Li, L

2004-02-03

The effects of various salts on the sol-gel transition of aqueous methylcellulose (MC) solutions have been studied systematically by means of a micro differential scanning calorimeter. It was found that the heating process was endothermic while the cooling process was exothermic for both MC solutions with and without salts. The addition of salts did not change the patterns of gelation and degelation of MC. However, the salts could shift the sol-gel transition and the gel-sol transition to lower or higher temperatures from a pure MC solution, depending on the salt type. These opposite effects were termed the salt-assisted and salt-suppressed sol-gel transitions. Either the salt-assisted transition or the salt-suppressed sol-gel transition was a function of salt concentration. In addition, each salt was found to have its own concentration limit for producing a stable aqueous solution of MC at a given concentration of MC, which was related to the anionic charge density of the salt. Cations were proved to have weaker effects than anions. The "salt-out strength", defined as the salt effect per mole of anion, was obtained for each anion studied. The thermodynamic mechanisms involved in the salt-assisted and salt-suppressed sol-gel transitions are discussed.

Experimental investigation of molten salt droplet quenching and solidification processes of heat recovery in thermochemical hydrogen production

International Nuclear Information System (INIS)

Ghandehariun, S.; Wang, Z.; Naterer, G.F.; Rosen, M.A.

2015-01-01

Highlights: • Thermal efficiency of a thermochemical cycle of hydrogen production is improved. • Direct contact heat recovery from molten salt is analyzed. • Falling droplets quenched into water are investigated experimentally. - Abstract: This paper investigates the heat transfer and X-ray diffraction patterns of solidified molten salt droplets in heat recovery processes of a thermochemical Cu–Cl cycle of hydrogen production. It is essential to recover the heat of the molten salt to enhance the overall thermal efficiency of the copper–chlorine cycle. A major portion of heat recovery within the cycle can be achieved by cooling and solidifying the molten salt exiting an oxygen reactor. Heat recovery from the molten salt is achieved by dispersing the molten stream into droplets. In this paper, an analytical study and experimental investigation of the thermal phenomena of a falling droplet quenched into water is presented, involving the droplet surface temperature during descent and resulting composition change in the quench process. The results show that it is feasible to quench the molten salt droplets for an efficient heat recovery process without introducing any material imbalance for the overall cycle integration.

Development of High Throughput Salt Separation System with Integrated Liquid Salt Separation - Salt Distillation Assembly

Energy Technology Data Exchange (ETDEWEB)

Kwon, Sangwoon; Park, K. M.; Kim, J. G.; Jeong, J. H.; Lee, S. J.; Park, S. B.; Kim, S. S.

2013-01-15

The capacity of a salt distiller should be sufficiently large to reach the throughput of uranium electro-refining process. In this study, an assembly composing a liquid separation sieve and a distillation crucible was developed for the sequential operation of a liquid salt separation and a vacuum distillation in the same tower. The feasibility of the sequential salt separation was examined by the rotation test of the sieve-crucible assembly and sequential operation of a liquid salt separation and a vacuum distillation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation. From the results of this study, it could be concluded that efficient salt separation can be realized by the sequential operation of liquid salt separation and vacuum distillation in one distillation tower since the operation procedures are simplified and no extra operation of cooling and reheating is necessary.

Effects of Gamma Irradiation on Quality in the Processing of Low Salted and Fermented Shrimp

International Nuclear Information System (INIS)

Shin Myung-Gon; Lee Cherl-Ho

2000-01-01

Irradiation technology was applied to develop low salted and fermented shrimp that has better sensory quality and a longer shelf-life without any food additives. Different levels of salt (10, 15, and 20%, w/w) were added to the salted and fermented shrimp and the samples were irradiated at 0, 2.5, 5.0, 7.5, and 10.0 kGy with a gamma source (Co-60). Proximate composition, salinity, water activity (a), pH, total bacterial count, and general acceptance were analyzed during fermentation at 15 degrees after irradiation. The proximate analysis, salinity, and a were not affected by gamma irradiation during fermentation. However, pH and total bacteria, as well as sensory evaluation, were changed variously with processing conditions such as sodium chloride concentration and irradiation dose. The combinations of 15% salt concentration with 10 kGy irradiation dose and 20% with 5 kGy or above were effective for shelf-life enhancement of the salted and fermented shrimp by adequate suppression of microorganisms during fermentation at 15 degrees. The results showed that the sensory quality of the sample was maintained up to 10 weeks after fermentation. Therefore, it was considered that gamma irradiation was effective in processing low salted and fermented shrimp and extending their shelf-life without adding any food additives

Synthesis and Characterization of Processable Polyaniline Salts

International Nuclear Information System (INIS)

Gul, Salma; Bilal, Salma; Shah, Anwar-ul-Haq Ali

2013-01-01

Polyaniline (PANI) is one of the most promising candidates for possible technological applications. PANI has potential applications in batteries, anion exchanger, tissue engineering, inhibition of steel corrosion, fuel cell, sensors and so on. However, its insolubility in common organic solvents limits its range of applications. In the present study an attempt has been made to synthesize soluble polyaniline salt via inverse polymerization pathway using benzoyl peroxide as oxidant and dodecylbenzenesulfonic acid (DBSA) as dopant as well as a surfactant. A mixture of chloroform and 2-butanol was used as dispersion medium for the first time. The influence of synthesis parameters such as concentration of aniline, benzoyl peroxide and DBSA on the yield and other properties of the resulting PANI salt was studied. The synthesized PANI salt was found to be completely soluble in DMSO, DMF, chloroform and in a mixture of toluene and 2-propanol. The synthesized polymer salt was also characterized with cyclic voltam-metry, SEM, XRD, UV-Vis spectroscopy and viscosity measurements. TGA was used to analyze the thermal properties of synthesized polymer. The extent of doping of the PANI salt was determined from UV-Vis spectra and TGA analysis. The activation energy for the degradation of the polymer was calculated with the help of TGA.

Salt acclimation process: a comparison between a sensitive and a tolerant Olea europaea cultivar.

Science.gov (United States)

Pandolfi, Camilla; Bazihizina, Nadia; Giordano, Cristiana; Mancuso, Stefano; Azzarello, Elisa

2017-03-01

Saline soils are highly heterogeneous in time and space, and this is a critical factor influencing plant physiology and productivity. Temporal changes in soil salinity can alter plant responses to salinity, and pre-treating plants with low NaCl concentrations has been found to substantially increase salt tolerance in different species in a process called acclimation. However, it still remains unclear whether this process is common to all plants or is only expressed in certain genotypes. We addressed this question by assessing the physiological changes to 100 mM NaCl in two contrasting olive cultivars (the salt-sensitive Leccino and the salt-tolerant Frantoio), following a 1-month acclimation period with 5 or 25 mM NaCl. The acclimation improved salt tolerance in both cultivars, but activated substantially different physiological adjustments in the tolerant and the sensitive cultivars. In the tolerant Frantoio the acclimation with 5 mM NaCl was more effective in increasing plant salt tolerance, with a 47% increase in total plant dry mass compared with non-acclimated saline plants. This enhanced biomass accumulation was associated with a 50% increase in K+ retention ability in roots. On the other hand, in the sensitive Leccino, although the acclimation process did not improve performance in terms of plant growth, pre-treatment with 5 and 25 mM NaCl substantially decreased salt-induced leaf cell ultrastructural changes, with leaf cell relatively similar to those of control plants. Taken together these results suggest that in the tolerant cultivar the acclimation took place primarily in the root tissues, while in the sensitive they occurred mainly at the shoot level. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Process to separate alkali metal salts from alkali metal reacted hydrocarbons

Energy Technology Data Exchange (ETDEWEB)

Gordon, John Howard; Alvare, Javier; Larsen, Dennis; Killpack, Jeff

2017-06-27

A process to facilitate gravimetric separation of alkali metal salts, such as alkali metal sulfides and polysulfides, from alkali metal reacted hydrocarbons. The disclosed process is part of a method of upgrading a hydrocarbon feedstock by removing heteroatoms and/or one or more heavy metals from the hydrocarbon feedstock composition. This method reacts the oil feedstock with an alkali metal and an upgradant hydrocarbon. The alkali metal reacts with a portion of the heteroatoms and/or one or more heavy metals to form an inorganic phase containing alkali metal salts and reduced heavy metals, and an upgraded hydrocarbon feedstock. The inorganic phase may be gravimetrically separated from the upgraded hydrocarbon feedstock after mixing at a temperature between about 350.degree. C. to 400.degree. C. for a time period between about 15 minutes and 2 hours.

ELECTRODIALYSIS IN THE CONVERSION STEP OF THE CONCENTRATED SALT SOLUTIONS IN THE PROCESS OF BATTERY SCRAP

Directory of Open Access Journals (Sweden)

S. I. Niftaliev

2014-01-01

Full Text Available Summary. The concentrated sodium sulfate solution is formed during the processing of waste battery scrap. A promising way to further treatment of the concentrated salt solution could be the conversion of these salts into acid and bases by electrodialysis, that can be reused in the same technical process cycle. For carrying out the process of conversion of salts into the corresponding acid and base several cells schemes with different combinations of cation, anion and bipolar membranes are used. At this article a comparative analysis of these cells is carried out. In the cells there were used the membranes МC-40, МА-41 and МB-2I. Acid and base solutions with higher concentration may be obtained during the process of electrodialysis in the circulation mode, when a predetermined amount of salt in the closed loop is run through a set of membranes to obtain the desired concentration of the product. The disadvantages of this method are the high cost of buffer tanks and the need to work with small volumes of treated solutions. In industrial applications it is advisable to use continuous electrodialysis with bipolar membranes, since this configuration allows to increase the number of repeating sections, which is necessary to reduce the energy costs. The increase of the removal rate of salts can be achieved by increasing the process steps, and to produce a more concentrated products after the conversion step can be applied electrodialysis-concentrator or evaporator.

A NOVEL PROCESS TO USE SALT CAVERNS TO RECEIVE SHIP BORNE LNG

Energy Technology Data Exchange (ETDEWEB)

Michael M. McCall; William M. Bishop; Marcus Krekel; James F. Davis; D. Braxton Scherz

2005-05-31

This cooperative research project validates use of man made salt caverns to receive and store the cargoes of LNG ships in lieu of large liquid LNG tanks. Salt caverns will not tolerate direct injection of LNG because it is a cryogenic liquid, too cold for contact with salt. This research confirmed the technical processes and the economic benefits of pressuring the LNG up to dense phase, warming it to salt compatible temperatures and then directly injecting the dense phase gas into salt caverns for storage. The use of salt caverns to store natural gas sourced from LNG imports, particularly when located offshore, provides a highly secure, large scale and lower cost import facility as an alternative to tank based LNG import terminals. This design can unload a ship in the same time as unloading at a tank based terminal. The Strategic Petroleum Reserve uses man made salt caverns to securely store large quantities of crude oil. Similarly, this project describes a novel application of salt cavern gas storage technologies used for the first time in conjunction with LNG receiving. The energy industry uses man made salt caverns to store an array of gases and liquids but has never used man made salt caverns directly in the importation of LNG. This project has adapted and expanded the field of salt cavern storage technology and combined it with novel equipment and processes to accommodate LNG importation. The salt cavern based LNG receiving terminal described in the project can be located onshore or offshore, but the focus of the design and cost estimates has been on an offshore location, away from congested channels and ports. The salt cavern based terminal can provide large volumes of gas storage, high deliverability from storage, and is simplified in operation compared to tank based LNG terminals. Phase I of this project included mathematical modeling that proved a salt cavern based receiving terminal could be built at lower capital cost, and would have significantly higher

Electrochemical extraction of samarium from molten chlorides in pyrochemical processes

International Nuclear Information System (INIS)

Castrillejo, Y.; Fernandez, P.; Medina, J.; Hernandez, P.; Barrado, E.

2011-01-01

This work concerns the electrochemical extraction of samarium from molten chlorides. In this way, the electrochemical behaviour of samarium ions has been investigated in the eutectic LiCl-KCl at the surface of tungsten, aluminium and aluminium coated tungsten electrodes. On a W inert electrode the electro-reduction of Sm(III) takes place in only one soluble-soluble electrochemical step Sm(III)/Sm(II). The electrochemical system Sm(II)/Sm(0) has not been observed within the electrochemical window, because of the prior reduction of Li(I) ions from the solvent, which inhibits the electro-extraction of Sm species from the salt on such a substrate. Sm metal in contact with the melt react to give Li(0) according to the reaction: Sm(0) + 2Li(I) ↔ Sm(II) + 2Li(0). On the contrary, on reactive Al electrodes the electrochemical system Sm(II)/Sm(0) was observed within the electroactive range. The potential shift of the redox couple is caused by the decrease of Sm activity in the metal phase due to the formation of Sm-Al alloys at the interface. The formation mechanism of the intermetallic compounds was studied in a melt containing: (i) both Sm(III) and Al(III) ions, using W and Al coated tungsten electrodes, and (ii) Sm(III) ions using an Al electrode. Analysis of the samples after potentiostatic electrolysis by X-ray diffraction and scanning electron microscopy (SEM) with energy dispersive X-ray spectroscopy (EDS), allowed the identification of Al 3 Sm and Al 2 Sm.

Electrochemical extraction of samarium from molten chlorides in pyrochemical processes

Energy Technology Data Exchange (ETDEWEB)

Castrillejo, Y., E-mail: ycastril@qa.uva.es [QUIANE/Dept Quimica Analitica, F. de Ciencias, Universidad de Valladolid, Prado de la Magdalena s/n, 47005 Valladolid (Spain); Fernandez, P. [QUIANE/Dept Quimica Analitica, F. de Ciencias, Universidad de Valladolid, Prado de la Magdalena s/n, 47005 Valladolid (Spain); Medina, J. [Dept Fisica Materia Condensada Cristalografia y Mineralogia, F. de Ciencias, Universidad de Valladolid, Prado de la Magdalena s/n, 47005 Valladolid (Spain); Hernandez, P. [Centro de Investigaciones Quimicas, Universidad Autonoma del Estado de Hidalgo, Carr. Pachuca-Tulancingo Km. 4.5, C.P. 42076 Pachuca, Hidalgo (Mexico); Barrado, E. [QUIANE/Dept Quimica Analitica, F. de Ciencias, Universidad de Valladolid, Prado de la Magdalena s/n, 47005 Valladolid (Spain)

2011-10-01

This work concerns the electrochemical extraction of samarium from molten chlorides. In this way, the electrochemical behaviour of samarium ions has been investigated in the eutectic LiCl-KCl at the surface of tungsten, aluminium and aluminium coated tungsten electrodes. On a W inert electrode the electro-reduction of Sm(III) takes place in only one soluble-soluble electrochemical step Sm(III)/Sm(II). The electrochemical system Sm(II)/Sm(0) has not been observed within the electrochemical window, because of the prior reduction of Li(I) ions from the solvent, which inhibits the electro-extraction of Sm species from the salt on such a substrate. Sm metal in contact with the melt react to give Li(0) according to the reaction: Sm(0) + 2Li(I) {r_reversible} Sm(II) + 2Li(0). On the contrary, on reactive Al electrodes the electrochemical system Sm(II)/Sm(0) was observed within the electroactive range. The potential shift of the redox couple is caused by the decrease of Sm activity in the metal phase due to the formation of Sm-Al alloys at the interface. The formation mechanism of the intermetallic compounds was studied in a melt containing: (i) both Sm(III) and Al(III) ions, using W and Al coated tungsten electrodes, and (ii) Sm(III) ions using an Al electrode. Analysis of the samples after potentiostatic electrolysis by X-ray diffraction and scanning electron microscopy (SEM) with energy dispersive X-ray spectroscopy (EDS), allowed the identification of Al{sub 3}Sm and Al{sub 2}Sm.

Processes and parameters involved in modeling radionuclide transport from bedded salt repositories. Final report. Technical memorandum

International Nuclear Information System (INIS)

Evenson, D.E.; Prickett, T.A.; Showalter, P.A.

1979-07-01

The parameters necessary to model radionuclide transport in salt beds are identified and described. A proposed plan for disposal of the radioactive wastes generated by nuclear power plants is to store waste canisters in repository sites contained in stable salt formations approximately 600 meters below the ground surface. Among the principal radioactive wastes contained in these canisters will be radioactive isotopes of neptunium, americium, uranium, and plutonium along with many highly radioactive fission products. A concern with this form of waste disposal is the possibility of ground-water flow occurring in the salt beds and endangering water supplies and the public health. Specifically, the research investigated the processes involved in the movement of radioactive wastes from the repository site by groundwater flow. Since the radioactive waste canisters also generate heat, temperature is an important factor. Among the processes affecting movement of radioactive wastes from a repository site in a salt bed are thermal conduction, groundwater movement, ion exchange, radioactive decay, dissolution and precipitation of salt, dispersion and diffusion, adsorption, and thermomigration. In addition, structural changes in the salt beds as a result of temperature changes are important. Based upon the half-lives of the radioactive wastes, he period of concern is on the order of a million years. As a result, major geologic phenomena that could affect both the salt bed and groundwater flow in the salt beds was considered. These phenomena include items such as volcanism, faulting, erosion, glaciation, and the impact of meteorites. CDM reviewed all of the critical processes involved in regional groundwater movement of radioactive wastes and identified and described the parameters that must be included to mathematically model their behavior. In addition, CDM briefly reviewed available echniques to measure these parameters

Process Evaluation and Costing of a Multifaceted Population-Wide Intervention to Reduce Salt Consumption in Fiji.

Science.gov (United States)

Webster, Jacqui; Pillay, Arti; Suku, Arleen; Gohil, Paayal; Santos, Joseph Alvin; Schultz, Jimaima; Wate, Jillian; Trieu, Kathy; Hope, Silvia; Snowdon, Wendy; Moodie, Marj; Jan, Stephen; Bell, Colin

2018-01-30

This paper reports the process evaluation and costing of a national salt reduction intervention in Fiji. The population-wide intervention included engaging food industry to reduce salt in foods, strategic health communication and a hospital program. The evaluation showed a 1.4 g/day drop in salt intake from the 11.7 g/day at baseline; however, this was not statistically significant. To better understand intervention implementation, we collated data to assess intervention fidelity, reach, context and costs. Government and management changes affected intervention implementation, meaning fidelity was relatively low. There was no active mechanism for ensuring food companies adhered to the voluntary salt reduction targets. Communication activities had wide reach but most activities were one-off, meaning the overall dose was low and impact on behavior limited. Intervention costs were moderate (FJD $277,410 or $0.31 per person) but the strategy relied on multi-sector action which was not fully operationalised. The cyclone also delayed monitoring and likely impacted the results. However, 73% of people surveyed had heard about the campaign and salt reduction policies have been mainstreamed into government programs. Longer-term monitoring of salt intake is planned through future surveys and lessons from this process evaluation will be used to inform future strategies in the Pacific Islands and globally.

Robot development for nuclear material processing

International Nuclear Information System (INIS)

Pedrotti, L.R.; Armantrout, G.A.; Allen, D.C.; Sievers, R.H. Sr.

1991-07-01

The Department of Energy is seeking to modernize its special nuclear material (SNM) production facilities and concurrently reduce radiation exposures and process and incidental radioactive waste generated. As part of this program, Lawrence Livermore National Laboratory (LLNL) lead team is developing and adapting generic and specific applications of commercial robotic technologies to SNM pyrochemical processing and other operations. A working gantry robot within a sealed processing glove box and a telerobot control test bed are manifestations of this effort. This paper describes the development challenges and progress in adapting processing, robotic, and nuclear safety technologies to the application. 3 figs

Vacuum distillation for the separation of LiCl-KCl eutectic salt and cadmium in pyro process

International Nuclear Information System (INIS)

Kwon, S. W.; Park, K. M.; Lee, S. J.; Park, S. B.; Woo, M. S.; Kim, K. R.; Kim, J. G.; Ahn, D. H.; Lee, H. S.

2010-10-01

Electrorefining is a key step in pyro processing. Electrorefining process is generally composed of two recovery steps- a deposit of uranium onto a solid cathode (electrorefining) and then the recovery of the remaining uranium and Tru (Transuranic) elements simultaneously by a liquid cadmium cathode (electrowinning). In this study, distillation experiments of a LiCl-KCl eutectic salt and cadmium metal were carried out to examine the distillation behaviors for the development of the electrorefining and the electrowinning processes. The experimental set-up was composed of a distillation tower with an evaporator and a condenser, vacuum pump, control unit, and an off gas treatment system. The solid-liquid separation prior to distillation of the LiCl-KCl eutectic salt was proposed and found to be feasible for the reduction of the burden of the distillation process. The LiCl-KCl eutectic salt was successfully distilled after the liquid salt separation. Distillation experiments for cadmium metal were also carried out. The apparent evaporation rates of LiCl-KCl eutectic salt and cadmium increased with an increasing temperature. The evaporation behaviors of cadmium metal and cadmium-cerium alloy were compared. Cadmium in the alloy was successfully distilled and separated from cerium. The evaporation rate of cadmium in the alloy was lower than that of cadmium metal. The low evaporation rate of the alloy was probably caused by the formation of an intermetallic compound and the residual salt during the preparation of the alloy. Therefore, the distillation temperature for the distillation of the liquid cathode should be higher than the distillation of cadmium metal. The measured evaporation rates of the eutectic salt and cadmium were compared with the values calculated by a relation based on the kinetics of gases. The theoretical values of the evaporation rate calculated by the Hertz-Langmuir relation were higher than the experimental values. The deviations were compensated for

Testing of Air Pulse Agitators to Support Design of Savannah River Site Highly Radioactive Processing at the Salt Waste Processing Facility

International Nuclear Information System (INIS)

Gallego, R.M.; Stephens, A.B.; Wilkinson, R.H.; Dev, H.; Suggs, P.C.

2006-01-01

The Salt Waste Processing Facility (SWPF) is intended to concentrate the highly radioactive constituents from waste salt solutions at the Savannah River Site (SRS). Air Pulse Agitators (APAs) were selected for process mixing in high-radiation locations at the SWPF. This technology has the advantage of no moving parts within the hot cell, eliminating potential failure modes and the need for maintenance within the high-radiation environment. This paper describes the results of APA tests performed to gain operational and performance data for the SWPF design. (authors)

Partitioning of actinides and fission products using molten salt electrorefining process

International Nuclear Information System (INIS)

Barbero, Jose A.; Wiesztort, Andres; Azcona, Alejandra; Bollini, Edgardo; Forchetti, Alberto; Orce, Alan

1999-01-01

Electrorefining is the key step of pyrometallurgical processing for separating actinides from fission products. In this work, the electrorefining process is carried out in a electrorefining cell that contains molten salts (49% LiCl- 51% KCL) floating on a liquid cadmium. The cell is operated under an inert atmosphere at 500 degree C. In this work we describe in detail the construction of the cell and the way of operation

Molten salt oxidation of organic hazardous waste with high salt content.

Science.gov (United States)

Lin, Chengqian; Chi, Yong; Jin, Yuqi; Jiang, Xuguang; Buekens, Alfons; Zhang, Qi; Chen, Jian

2018-02-01

Organic hazardous waste often contains some salt, owing to the widespread use of alkali salts during industrial manufacturing processes. These salts cause complications during the treatment of this type of waste. Molten salt oxidation is a flameless, robust thermal process, with inherent capability of destroying the organic constituents of wastes, while retaining the inorganic ingredients in the molten salt. In the present study, molten salt oxidation is employed for treating a typical organic hazardous waste with a high content of alkali salts. The hazardous waste derives from the production of thiotriazinone. Molten salt oxidation experiments have been conducted using a lab-scale molten salt oxidation reactor, and the emissions of CO, NO, SO 2 , HCl and dioxins are studied. Impacts are investigated from the composition of the molten salts, the types of feeding tube, the temperature of molten carbonates and the air factor. Results show that the waste can be oxidised effectively in a molten salt bath. Temperature of molten carbonates plays the most important role. With the temperature rising from 600 °C to 750 °C, the oxidation efficiency increases from 91.1% to 98.3%. Compared with the temperature, air factor has but a minor effect, as well as the composition of the molten salts and the type of feeding tube. The molten carbonates retain chlorine with an efficiency higher than 99.9% and the emissions of dioxins are below 8 pg TEQ g -1 sample. The present study shows that molten salt oxidation is a promising alternative for the disposal of organic hazardous wastes containing a high salt content.

Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

International Nuclear Information System (INIS)

CHANG, ROBERT

2006-01-01

All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

International Nuclear Information System (INIS)

Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

1992-01-01

This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

Plutonium and americium recovery from spent molten-salt-extraction salts with aluminum-magnesium alloys

International Nuclear Information System (INIS)

Cusick, M.J.; Sherwood, W.G.; Fitzpatrick, R.F.

1984-01-01

Development work was performed to determine the feasibility of removing plutonium and americium from spent molten-salt-extraction (MSE) salts using Al-Mg alloys. If the product buttons from this process are compatible with subsequent aqueous processing, the complex chloride-to-nitrate aqueous conversion step which is presently required for these salts may be eliminated. The optimum alloy composition used to treat spent 8 wt % MSE salts in the past yielded poor phase-disengagement characteristics when applied to 30 mol % salts. After a limited investigation of other alloy compositions in the Al-Mg-Pu-Am system, it was determined that the Al-Pu-Am system could yield a compatible alloy. In this system, experiments were performed to investigate the effects of plutonium loading in the alloy, excess magnesium, age of the spent salt on actinide recovery, phase disengagement, and button homogeneity. Experimental results indicate that 95 percent plutonium recoveries can be attained for fresh salts. Further development is required for backlog salts generated prior to 1981. A homogeneous product alloy, as required for aqueous processing, could not be produced

Characterization of the molten salt reactor experiment fuel and flush salts

International Nuclear Information System (INIS)

Williams, D.F.; Peretz, F.J.

1996-01-01

Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

High salt diet induces metabolic alterations in multiple biological processes of Dahl salt-sensitive rats.

Science.gov (United States)

Wang, Yanjun; Liu, Xiangyang; Zhang, Chen; Wang, Zhengjun

2018-06-01

High salt induced renal disease is a condition resulting from the interactions of genetic and dietary factors causing multiple complications. To understand the metabolic alterations associated with renal disease, we comprehensively analyzed the metabonomic changes induced by high salt intake in Dahl salt-sensitive (SS) rats using GC-MS technology and biochemical analyses. Physiological features, serum chemistry, and histopathological data were obtained as complementary information. Our results showed that high salt (HS) intake for 16 weeks caused significant metabolic alterations in both the renal medulla and cortex involving a variety pathways involved in the metabolism of organic acids, amino acids, fatty acids, and purines. In addition, HS enhanced glycolysis (hexokinase, phosphofructokinase and pyruvate kinase) and amino acid metabolism and suppressed the TCA (citrate synthase and aconitase) cycle. Finally, HS intake caused up-regulation of the pentose phosphate pathway (glucose 6-phosphate dehydrogenase and 6-phosphogluconate dehydrogenase), the ratio of NADPH/NADP + , NADPH oxidase activity and ROS production, suggesting that increased oxidative stress was associated with an altered PPP pathway. The metabolic pathways identified may serve as potential targets for the treatment of renal damage. Our findings provide comprehensive biochemical details about the metabolic responses to a high salt diet, which may contribute to the understanding of renal disease and salt-induced hypertension in SS rats. Copyright © 2018. Published by Elsevier Inc.

An introduction to pyrochemistry with emphasis on nuclear applications

International Nuclear Information System (INIS)

Johnson, I.

2001-01-01

This report illustrates the application of thermodynamic principles to further understanding of pyrochemical processes useful for the recovery and purification of nuclear materials. It draws extensively upon the theoretical and experimental work conducted over nearly four decades in the Chemical Technology Division (formerly called the Chemical Engineering Division) of Argonne National Laboratory. After a brief discussion of fundamental thermodynamic principles and equations, they are applied to the relative distribution of actinides and other elements in molten salts and liquid metal solutions and to solubilities in liquid metal solutions

Doping Polymer Semiconductors by Organic Salts: Toward High-Performance Solution-Processed Organic Field-Effect Transistors.

Science.gov (United States)

Hu, Yuanyuan; Rengert, Zachary D; McDowell, Caitlin; Ford, Michael J; Wang, Ming; Karki, Akchheta; Lill, Alexander T; Bazan, Guillermo C; Nguyen, Thuc-Quyen

2018-04-24

Solution-processed organic field-effect transistors (OFETs) were fabricated with the addition of an organic salt, trityl tetrakis(pentafluorophenyl)borate (TrTPFB), into thin films of donor-acceptor copolymer semiconductors. The performance of OFETs is significantly enhanced after the organic salt is incorporated. TrTPFB is confirmed to p-dope the organic semiconductors used in this study, and the doping efficiency as well as doping physics was investigated. In addition, systematic electrical and structural characterizations reveal how the doping enhances the performance of OFETs. Furthermore, it is shown that this organic salt doping method is feasible for both p- and n-doping by using different organic salts and, thus, can be utilized to achieve high-performance OFETs and organic complementary circuits.

Engineering development studies for molten-salt breeder reactor processing No. 22

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1976-06-01

Processing methods are being developed for use in a close-coupled facility for removing fission products, corrosion products, and fissile materials from the MSBR fuel. This report discusses the autoresistance heating for the continuous fluorinator, the metal transfer experiment, experiments for the salt-metal contactor, and fuel reconstitution. 10 fig

Process Evaluation and Costing of a Multifaceted Population-Wide Intervention to Reduce Salt Consumption in Fiji

Directory of Open Access Journals (Sweden)

Jacqui Webster

2018-01-01

Full Text Available This paper reports the process evaluation and costing of a national salt reduction intervention in Fiji. The population-wide intervention included engaging food industry to reduce salt in foods, strategic health communication and a hospital program. The evaluation showed a 1.4 g/day drop in salt intake from the 11.7 g/day at baseline; however, this was not statistically significant. To better understand intervention implementation, we collated data to assess intervention fidelity, reach, context and costs. Government and management changes affected intervention implementation, meaning fidelity was relatively low. There was no active mechanism for ensuring food companies adhered to the voluntary salt reduction targets. Communication activities had wide reach but most activities were one-off, meaning the overall dose was low and impact on behavior limited. Intervention costs were moderate (FJD $277,410 or $0.31 per person but the strategy relied on multi-sector action which was not fully operationalised. The cyclone also delayed monitoring and likely impacted the results. However, 73% of people surveyed had heard about the campaign and salt reduction policies have been mainstreamed into government programs. Longer-term monitoring of salt intake is planned through future surveys and lessons from this process evaluation will be used to inform future strategies in the Pacific Islands and globally.

Ingot formation using uranium dendrites recovered by electrolysis in LiCl-KCl-PuCl3-UCl3 melt

International Nuclear Information System (INIS)

Mineo Fukushima; Akira Nakayoshi; Shinichi Kitawaki; Masaki Kurata; Noboru Yahagi

2008-01-01

Products on solid cathodes recovered by the metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions of uranium formation, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In these tests, it is confirmed that uranium ingots can be obtained with heating the products more than melting temperature of metal uranium under atmospheric pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Covered salt can be removed after ingot formation. Inside the ingot, there were a lump of uranium and dark brown colored dross was observed. Material balance of uranium is 77 ∼ 96%, that of plutonium is 71 ∼ 109%, and that of americium that is a volatile substance more than uranium and plutonium become 79 ∼ 119%. Volatilization of americium is very small under the condition of high temperature. (authors)

Ingot formation using uranium dendrites recovered by electrolysis in LiCl-KCl-PuCl{sub 3}-UCl{sub 3} melt

Energy Technology Data Exchange (ETDEWEB)

Mineo Fukushima; Akira Nakayoshi; Shinichi Kitawaki [Japan Atomic Energy Agency (JAEA), 4-33 Muramatsu Tokai-mura Naka-gun, Ibaraki, 319-1194 (Japan); Masaki Kurata; Noboru Yahagi [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1 Iwadokita Komae-shi, Tokyo, 201-8511 (Japan)

2008-07-01

Products on solid cathodes recovered by the metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions of uranium formation, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In these tests, it is confirmed that uranium ingots can be obtained with heating the products more than melting temperature of metal uranium under atmospheric pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Covered salt can be removed after ingot formation. Inside the ingot, there were a lump of uranium and dark brown colored dross was observed. Material balance of uranium is 77 {approx} 96%, that of plutonium is 71 {approx} 109%, and that of americium that is a volatile substance more than uranium and plutonium become 79 {approx} 119%. Volatilization of americium is very small under the condition of high temperature. (authors)

Defining a metal-based waste form for IFR pyroprocessing wastes

International Nuclear Information System (INIS)

McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.

1994-01-01

Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts

UHPLC-MS/MS Quantification Combined with Chemometrics for Comparative Analysis of Different Batches of Raw, Wine-Processed, and Salt-Processed Radix Achyranthis Bidentatae

Directory of Open Access Journals (Sweden)

Liu Yang

2018-03-01

Full Text Available An accurate and reliable method using ultra-high performance liquid chromatography combined with triple quadrupole tandem mass spectrometry (UHPLC–MS/MS was established for simultaneous quantification of five major bioactive analytes in raw, wine-processed, and salt-processed Radix Achyranthis bidentatae (RAB. The results showed that this method exhibited desirable sensitivity, precision, stability, and repeatability. The overall intra-day and inter-day variations (RSD were in the range of 1.57–2.46 and 1.51–3.00%, respectively. The overall recoveries were 98.58–101.48% with a relative standard deviation (RSD of 0.01–1.86%. In addition, the developed approach was applied to 21 batches of raw, wine-processed, and salt-processed samples of RAB. Hierarchical clustering analysis (HCA, principal component analysis (PCA, heat map, and boxplot analysis were performed to evaluate the quality of raw, wine-processed, and salt-processed RAB collected from different regions. The chemometrics combined with the quantitative analysis based on UHPLC–MS/MS results indicated that the content of five analytes increased significantly in processed RAB compared to raw RAB.

Research on the actinide chemistry in Nuclear Fuel Cycle

Energy Technology Data Exchange (ETDEWEB)

Song, Kyseok; Park, Yong Joon; Cho, Young Hwan; and others

2012-04-15

Fundamental technique to measure chemical behaviors and properties of lanthanide and actinide in radioactive waste is necessary for the development of pryochemical process. First stage, the electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipments, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media. In the second stage, measurement system for physical properties at pyrochemical process such as viscosity, melting point and conductivity is established, and property database at different compositions of lanthanide and actinide is collected. And, both interactions between elements and properties with different potential are measured at binary composition of actinide-lanthanide in molten salt using electrochemical/spectroscopic integrated measurement system.

Process-based management approaches for salt desert shrublands dominated by downy brome

Science.gov (United States)

Downy brome grass (Bromus tectorum L.) invasion has severely altered key ecological processes such as disturbance regimes, soil nutrient cycling, community assembly, and successional pathways in semi-arid Great Basin salt desert shrublands. Restoring the structure and function of these severly alte...

Salt Stability - The Effect of pHmax on Salt to Free Base Conversion.

Science.gov (United States)

Hsieh, Yi-Ling; Merritt, Jeremy M; Yu, Weili; Taylor, Lynne S

2015-09-01

The aim of this study was to investigate how the disproportionation process can be impacted by the properties of the salt, specifically pHmax. Five miconazole salts and four sertraline salts were selected for this study. The extent of conversion was quantified using Raman spectroscopy. A mathematical model was utilized to estimate the theoretical amount of conversion. A trend was observed that for a given series of salts of a particular basic compound (both sertraline and miconazole are bases), the extent of disproportionation increases as pHmax decreases. Miconazole phosphate monohydrate and sertraline mesylate, although exhibiting significantly different pHmax values (more than 2 units apart), underwent a similar extent of disproportionation, which may be attributed to the lower buffering capacity of sertraline salts. This work shows that the disproportionation tendency can be influenced by pHmax and buffering capacity and thus highlights the importance of selecting the appropriate salt form during the screening process in order to avoid salt-to-free form conversion.

Salt Separation from Uranium Deposits in Integrated Crucible

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non-volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation.

Water uptake by salts during the electrolyte processing for thermal batteries

Science.gov (United States)

Masset, Patrick; Poinso, Jean-Yves; Poignet, Jean-Claude

Water uptake of single salts and electrolytes were measured in industrial conditions (dry-room). The water uptake rate ϑ (g h -1 cm -2) was expressed with respect to the apparent area of contact of the salt with atmosphere of the dry room. The water uptake by potassium-based salts was very low. LiF and LiCl salts were found to behave similarly. For LiBr- and LiI-based salts and mixtures, we pointed out a linear relationship between the water uptake and the elapsed time. Water uptake by magnesium oxide reached a limit after 200 h. This work provides a set of data concerning the rate of water uptake by single salts, salt mixtures and magnesia used in thermal battery electrolytes.

Electrochemical behaviour of ThF4 in LiCl-KCl eutectic

International Nuclear Information System (INIS)

Agarwal, Renu; Mukerjee, S.K.; Ramakumar, K.L.

2016-01-01

Pyrochemical processing of nuclear fuels is an integral part of molten salt reactor technology. Thorium containing molten salt reactors are relevant for transmutation of minor actinides and utilizations of vast thorium resources. Understanding chemical and electrochemical behavior of actinide and lanthanide ions in chloride melts is an important task in this process. A eutectic of lithium and potassium chloride is found to be the most suitable salt for this study due to its low melting, high reduction potential and significant solubility of most of the actinide and lanthanide salts. However, electrochemistry of thorium in chloride melt is inconsistent due to dispute over the presence of Th(II) cation along with Th(IV). In the present experiments, cyclic-voltammograms of ThF 4 in LiCl-KCl eutectic salt were measured using three electrodes cell, where all the electrodes were 1 mm molybdenum wires and counter electrode wire was twisted into a spiral shape to increase surface area and was used for mixing the salts in molten state. Cyclic-voltammograms of pure eutectic melt and with ThF 4 were measured at scanning rates 50-300 mV/s, at 648 K and 723 K. Diffusion coefficient of Th 4+ was calculated using Delahay equation

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

Navratil, J.D.; Stewart, A.E.

1996-01-01

MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO2-Al2O3-P2O5(SAP) Inorganic Composite: Part 1. Dechlorination Behavior of LiCl-KCl and Characteristics of Consolidation

International Nuclear Information System (INIS)

Cho, In Hak; Park, Hwan Seo; Ahn, Soo Na; Kim, In Tae; Cho, Yong Zun

2012-01-01

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a SiO 2 -Al 2 O 3 -P 2 O 5 (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about 400 degree C for LiCl and the second was about 700 degree C for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a SiO 2 -Al 2 O 3 -P 2 O 5 (SAP) could be considered as one of alternatives for the immobilization of waste salt.

Glovebox design requirements for molten salt oxidation processing of transuranic waste

International Nuclear Information System (INIS)

Ramsey, K.B.; Acosta, S.V.; Wernly, K.D.

1998-01-01

This paper presents an overview of potential technologies for stabilization of 238 Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from 238 Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented

Feasibility of Th-U separation through a pyrochemical route in molten LiCl-KCl eutectic

International Nuclear Information System (INIS)

Pakhui, Gurudas; Ghosh, Suddhasattwa; Prabhakara Reddy, B.; Nagarajan, K.

2014-01-01

Molten salt electrorefining is a high temperature electrometallurgical process developed for the reprocessing of spent UPu-Zr fuel employing LiCl-KCl as the electrolyte. A possible application of this high temperature electrochemical process could be in the separation of U from Th matrix for Th based fuels. It therefore becomes important to investigate the reduction behaviour of Th 4+ in the molten salt and compare with that of U 3+ . The present study is on the electrochemical behaviour of Th 4+ in LiCl-KCl. Electrochemical studies were also carried out on the LiCl-KCl-UCl 3 -ThCl 4 system using transient electrochemical techniques. The cyclic voltammograms for Th 4+ /Th redox couple at various scan rates at 723 K is shown. Reduction of Th 4+ to Th was found to be quasi-reversible at lower scan rates and irreversible at higher scan rates using cyclic voltammetry and convolution voltammetry. The number of electrons transferred for the reduction process was calculated to be ∼ 4 using various techniques like cyclic voltammetry, chronopotentiometry and convolution voltammetry

Recovery of transuranics from process residues

International Nuclear Information System (INIS)

Gray, J.H.; Gray, L.W.

1987-01-01

Process residues are generated at both the Rocky Flats Plant (RFP) and the Savannah River Plant (SRP) during aqueous chemical and pyrochemical operations. Frequently, process operations will result in either impure products or produce residues sufficiently contaminated with transuranics to be nondiscardable as waste. Purification and recovery flowsheets for process residues have been developed to generate solutions compatible with subsequent Purex operations and either solid or liquid waste suitable for disposal. The ''scrub alloy'' and the ''anode heel alloy'' are examples of materials generated at RFP which have been processed at SRP using the developed recovery flowsheets. Examples of process residues being generated at SRP for which flowsheets are under development include LECO crucibles and alpha-contaminated hydraulic oil

Improving molten fluoride salt and Xe135 barrier property of nuclear graphite by phenolic resin impregnation process

Science.gov (United States)

He, Zhao; Lian, Pengfei; Song, Yan; Liu, Zhanjun; Song, Jinliang; Zhang, Junpeng; Feng, Jing; Yan, Xi; Guo, Quangui

2018-02-01

A densification process has been conducted on isostatic graphite (IG-110, TOYO TANSO CO., Ltd., Japan) by impregnating phenolic resin to get the densified isostatic graphite (D-IG-110) with pore diameter of nearly 11 nm specifically for molten salt reactor application. The microstructure, mechanical, thermophysical and other properties of graphite were systematically investigated and compared before and after the densification process. The molten fluoride salt and Xe135 penetration in the graphite were evaluated in a high-pressure reactor and a vacuum device, respectively. Results indicated that D-IG-110 exhibited improved properties including infiltration resistance to molten fluoride salt and Xe135 as compared to IG-110 due to its low porosity of 2.8%, the average pore diameter of 11 nm and even smaller open pores on the surface of the graphite. The fluoride salt infiltration amount of IG-110 was 13.5 wt% under 1.5 atm and tended to be saturated under 3 atm with the fluoride salt occupation of 14.8 wt%. As to the D-IG-110, no salts could be detected even up to 10 atm attempted loading. The helium diffusion coefficient of D-IG-110 was 6.92 × 10-8 cm2/s, significantly less than 1.21 × 10-2 cm2/s of IG-110. If these as-produced properties for impregnated D-IG-110 could be retained during MSR operation, the material could prove effective at inhibiting molten fluoride salt and Xe135 inventories in the graphite.

Processed foods as an integral part of universal salt iodization programs: a review of global experience and analyses of Bangladesh and Pakistan.

Science.gov (United States)

Spohrer, Rebecca; Garrett, Greg S; Timmer, Arnold; Sankar, Rajan; Kar, Basanta; Rasool, Faiz; Locatelli-Rossi, Lorenzo

2012-12-01

Despite the reference to salt for food processing in the original definition of universal salt iodization (USI), national USI programs often do not explicitly address food industry salt. This may affect program impact and sustainability, given the increasing consumption of processed foods in developing countries. To review experience of the use of iodized salt in the food industry globally, and analyze the market context in Bangladesh and Pakistan to test whether this experience may be applicable to inform improved national USI programming in developing countries. A review of relevant international experience was undertaken. In Bangladesh and Pakistan, local rural market surveys were carried out. In Bangladesh, structured face-to-face interviews with bakers and indepth interviews with processed food wholesalers and retailers were conducted. In Pakistan, face-to-face structured interviews were conducted with food retailers and food labels were checked. Experience from industrialized countries reveals impact resulting from the use of iodized salt in the food industry. In Bangladesh and Pakistan, bread, biscuits, and snacks containing salt are increasingly available in rural areas. In Bangladesh, the majority of bakers surveyed claimed to use iodized salt. In Pakistan, 6 of 362 unique product labels listed iodized salt. Successful experience from developed countries needs to be adapted to the developing country context. The increasing availability of processed foods in rural Bangladesh and Pakistan provides an opportunity to increase iodine intake. However, the impact of this intervention remains to be quantified. To develop better national USI programs, further data are required on processed food consumption across population groups, iodine contents of food products, and the contribution of processed foods to iodine nutrition.

EXAMINE AND EVALUATE A PROCESS TO USE SALT CAVERNS TO RECEIVE SHIP BORNE LIQUEFIED NATURAL GAS

Energy Technology Data Exchange (ETDEWEB)

Michael M. McCall; William M. Bishop; D. Braxton Scherz

2003-04-24

The goal of the U.S. Department of Energy cooperative research project is to define, describe, and validate, a process to utilize salt caverns to receive and store the cargoes of LNG ships. The project defines the process as receiving LNG from a ship, pumping the LNG up to cavern injection pressures, warming it to cavern compatible temperatures, injecting the warmed vapor directly into salt caverns for storage, and distribution to the pipeline network. The performance of work under this agreement is based on U.S. Patent 5,511,905, and other U.S. and Foreign pending patent applications. The cost sharing participants in the research are The National Energy Technology Laboratory (U.S. Department of Energy), BP America Production Company, Bluewater Offshore Production Systems (U.S.A.), Inc., and HNG Storage, L.P. Initial results indicate that a salt cavern based receiving terminal could be built at about half the capital cost, less than half the operating costs and would have significantly higher delivery capacity, shorter construction time, and be much more secure than a conventional liquid tank based terminal. There is a significant body of knowledge and practice concerning natural gas storage in salt caverns, and there is a considerable body of knowledge and practice in handling LNG, but there has never been any attempt to develop a process whereby the two technologies can be combined. Salt cavern storage is infinitely more secure than surface storage tanks, far less susceptible to accidents or terrorist acts, and much more acceptable to the community. The project team developed conceptual designs of two salt cavern based LNG terminals, one with caverns located in Calcasieu Parish Louisiana, and the second in Vermilion block 179 about 50 miles offshore Louisiana. These conceptual designs were compared to conventional tank based LNG terminals and demonstrate superior security, economy and capacity. The potential for the development of LNG receiving terminals

Glovebox design requirements for molten salt oxidation processing of transuranic waste

Energy Technology Data Exchange (ETDEWEB)

Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

1998-12-31

This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

Use of sodium salt electrolysis in the process of continuous ...

Indian Academy of Sciences (India)

This paper presents test results concerning the selection of sodium salt for the technology of continuous modification of the EN AC-AlSi12 alloy, which is based on electrolysis of sodium salts, occurring directly in a crucible with liquid alloy. Sodium ions formed as a result of the sodium salt dissociation and the electrolysis are ...

Hydrothermal processing of fermentation residues in a continuous multistage rig – Operational challenges for liquefaction, salt separation, and catalytic gasification

International Nuclear Information System (INIS)

Zöhrer, H.; De Boni, E.; Vogel, F.

2014-01-01

Fermentation residues are a waste stream of biomethane production containing substantial amounts of organic matter, and thus representing a primary energy source which is mostly unused. For the first time this feedstock was tested for catalytic gasification in supercritical water (T ≥ 374 °C, p ≥ 22 MPa) for methane production. The processing steps include hydrothermal liquefaction, salt separation, as well as catalytic gasification over a ruthenium catalyst in supercritical water. In continuous experiments at a feed rate of 1 kg h −1 a partial liquefaction and carbonization of some of the solids was observed. Significant amounts of heavy tars were formed. Around 50% of the feed carbon remained in the rig. Furthermore, a homogeneous coke was formed, presumably originating from condensed tars. The mineralization of sulfur and its separation in the salt separator was insufficient, because most of the sulfur was still organically bound after liquefaction. Desalination was observed at a salt separator set point temperature of 450 °C and 28 MPa; however, some of the salts could not be withdrawn as a concentrated brine. At 430 °C no salt separation took place. Higher temperatures in the salt separator were found to promote tar and coke formation, resulting in conflicting process requirements for efficient biomass liquefaction and desalination. In the salt separator effluent, solid crystals identified as struvite (magnesium ammonium phosphate) were found. This is the first report of struvite formation from a supercritical water biomass conversion process and represents an important finding for producing a fertilizer from the separated salt brine. - Highlights: • Continuous processing of fermentation residues in sub- and supercritical water. • Continuous separation of salt brines at supercritical water conditions. • Struvite crystals (magnesium ammonium phosphate) were recovered from the effluent. • Separation of sulfur from the biomass could

Mixing of zeolite powders and molten salt

International Nuclear Information System (INIS)

Pereira, C.; Zyryanov, V.N.; Lewis, M.A.; Ackerman, J.P.

1996-01-01

Transuranics and fission products in a molten salt can be incorporated into zeolite A by an ion exchange process and by a batch mixing or blending process. The zeolite is then mixed with glass and consolidated into a monolithic waste form for geologic disposal. Both processes require mixing of zeolite powders with molten salt at elevated temperatures (>700 K). Complete occlusion of salt and a uniform distribution of chloride and fission products are desired for incorporation of the powders into the final waste form. The relative effectiveness of the blending process was studied over a series of temperature, time, and composition profiles. The major criteria for determining the effectiveness of the mixing operations were the level and uniformity of residual free salt in the mixtures. High operating temperatures (>775 K) improved salt occlusion. Reducing the chloride levels in the mixture to below 80% of the full salt capacity of the zeolite significantly reduced the free salt level in the final product

Ultrasonic characterization of pork meat salting

International Nuclear Information System (INIS)

García-Pérez, J V; De Prados, M; Pérez-Muelas, N; Cárcel, J A; Benedito, J

2012-01-01

Salting process plays a key role in the preservation and quality of dry-cured meat products. Therefore, an adequate monitoring of salt content during salting is necessary to reach high quality products. Thus, the main objective of this work was to test the ability of low intensity ultrasound to monitor the salting process of pork meat. Cylindrical samples (diameter 36 mm, height 60±10 mm) of Biceps femoris were salted (brine 20% NaCl, w/w) at 2 °C for 1, 2, 4 and 7 days. During salting and at each experimental time, three cylinders were taken in order to measure the ultrasonic velocity at 2 °C. Afterwards, the cylinders were split in three sections (height 20 mm), measuring again the ultrasonic velocity and determining the salt and the moisture content by AOAC standards. In the whole cylinders, moisture content was reduced from 763 (g/kg sample) in fresh samples to 723 (g/kg sample) in samples salted for 7 days, while the maximum salt gain was 37.3 (g/kg sample). Although, moisture and salt contents up to 673 and 118 (g/kg sample) were reached in the sections of meat cylinders, respectively. During salting, the ultrasonic velocity increased due to salt gain and water loss. Thus, significant (p 2 = 0.975) and moisture (R 2 = 0.863) contents. In addition, the change of the ultrasonic velocity with the increase of the salt content showed a good agreement with the Kinsler equation. Therefore, low intensity ultrasound emerges as a potential technique to monitor, in a non destructive way, the meat salting processes carried out in the food industry.

Salts of alkali metal anions and process of preparing same

Science.gov (United States)

Dye, James L.; Ceraso, Joseph M.; Tehan, Frederick J.; Lok, Mei Tak

1978-01-01

Compounds of alkali metal anion salts of alkali metal cations in bicyclic polyoxadiamines are disclosed. The salts are prepared by contacting an excess of alkali metal with an alkali metal dissolving solution consisting of a bicyclic polyoxadiamine in a suitable solvent, and recovered by precipitation. The salts have a gold-color crystalline appearance and are stable in a vacuum at -10.degree. C. and below.

Molten-salt converter reactors

International Nuclear Information System (INIS)

Perry, A.M.

1975-01-01

Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

Distillation of LiCl from the LiCl-Li2O molten salt of the electrolytic reduction process

International Nuclear Information System (INIS)

Kim, I.S.; Oh, S.C.; Im, H.S.; Hur, J.M.; Lee, H.S.

2013-01-01

Electrolytic reduction of the uranium oxide in LiCl-Li 2 O molten salt for the treatment of spent nuclear fuel requires the separation of the residual salt from the reduced metal product, which contains about 20 wt% salt. In order to separate the residual salt and reuse it in the electrolytic reduction, a vacuum distillation process was developed. Lab-scale distillation equipment was designed and installed in an argon atmosphere glove box. The equipment consisted of an evaporator in which the reduced metal product was contained and exposed to a high temperature and reduced pressure; a receiver; and a vertically oriented condenser that operated at a temperature below the melting point of lithium chloride. We performed experiments with LiCl-Li 2 O salt to evaluate the evaporation rate of LiCl salt and varied the operating temperature to discern its effect on the behavior of salt evaporation. Complete removal of the LiCl salt from the evaporator was accomplished by reducing the internal pressure to <100 mTorr and heating to 900 deg C. We achieved evaporation efficiency as high as 100 %. (author)

Properties of glass-bonded zeolite monoliths

International Nuclear Information System (INIS)

Lewis, M.A.; Fischer, D.F.; Murphy, C.D.

1994-01-01

It has been shown that mineral waste forms can be used to immobilize waste salt generated during the pyrochemical processing of spent fuel from the Integral Fast Reactor (IFR). Solid, leach resistant monoliths were formed by hot-pressing mixtures of salt-occluded zeolite A powders and glass frit at 990 K and 28 MPa. Additional samples have now been fabricated and tested. Normalized release rates for all elements, including iodide and chloride, were less than 1 g/m 2 d in 28-day tests in deionized water and in brine at 363 K (90 degrees C). Preliminary results indicate that these rates fall with time with both leachants and that the zeolite phase in the glass-bonded zeolite does not function as an ion exchanger. Some material properties were measured. The Poisson ratio and Young's modulus were slightly smaller in glass-bonded zeolite than in borosilicate glass. Density depended on zeolite fraction. The glass-bonded zeolite represents a promising mineral waste form for IFR salt

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}(SAP) Inorganic Composite: Part 1. Dechlorination Behavior of LiCl-KCl and Characteristics of Consolidation

Energy Technology Data Exchange (ETDEWEB)

Cho, In Hak; Park, Hwan Seo; Ahn, Soo Na; Kim, In Tae; Cho, Yong Zun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-03-15

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5} (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about 400 degree C for LiCl and the second was about 700 degree C for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}(SAP) could be considered as one of alternatives for the immobilization of waste salt.

Development of pyrometallurgical partitioning technology of long-lived nuclides. Recovery of volatile chlorides for chlorination process using molten salt trap. 1

International Nuclear Information System (INIS)

Hijikata, Takatoshi; Nakamura, Kyosei; Kurata, Masateru; Konagaya, Hideaki

1997-01-01

The dry process for partitioning of long-lived nuclides from high level radioactive waste has been developed. One of the subjects for development of this process is the recovering of the volatilization of chlorides for the chlorination process. We proposed that the volatile chlorides were recovered by the molten salt trap. We researched the behavior of volatile chlorides (ferric chloride, zirconium tetra-chloride and molybdenum pent-chloride) in LiCl-KCl eutectic salt. In this result, the volatile rate of these chlorides was slower than the volatile rate of undissolved chlorides in LiCl-KCl eutectic salt. Also, we make a prototype of molten salt trap for recovering the volatile chlorides and tested the performance of this experimental apparatus and recovering ratio of volatile chlorides. This trap has a good performance of recovering volatile chlorides. (author)

Salt Removal from the Uranium Deposits of Electrorefiner

Energy Technology Data Exchange (ETDEWEB)

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

2010-10-15

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

Salt Removal from the Uranium Deposits of Electrorefiner

International Nuclear Information System (INIS)

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

2010-01-01

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

OPERATIONS REVIEW OF THE SAVANNAH RIVER SITE INTEGRATED SALT DISPOSITION PROCESS - 11327

Energy Technology Data Exchange (ETDEWEB)

Peters, T.; Poirier, M.; Fondeur, F.; Fink, S.; Brown, S.; Geeting, M.

2011-02-07

The Savannah River Site (SRS) is removing liquid radioactive waste from its Tank Farm. To treat waste streams that are low in Cs-137, Sr-90, and actinides, SRS developed the Actinide Removal Process and implemented the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The Actinide Removal Process contacts salt solution with monosodium titanate to sorb strontium and select actinides. After monosodium titanate contact, the resulting slurry is filtered to remove the monosodium titanate (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), and strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to the Defense Waste Processing Facility for vitrification. The facilities began radiological processing in April 2008 and started processing of the third campaign ('MarcoBatch 3') of waste in June 2010. Campaigns to date have processed {approx}1.2 million gallons of dissolved saltcake. Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples for each waste batch prior to processing. Testing included monosodium titanate sorption of strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass transfer. This paper describes the tests conducted and compares results from facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Additionally, the poster describes lessons learned during

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

Science.gov (United States)

Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

2005-12-01

Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

Development of an integrated crucible for the salt separation

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-15

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

Significant Modules and Biological Processes between Active Components of Salvia miltiorrhiza Depside Salt and Aspirin

Directory of Open Access Journals (Sweden)

Yuan Li

2016-01-01

Full Text Available The aim of this study is to examine and compare the similarities and differences between active components of S. miltiorrhiza depside salt and aspirin using perspective of pharmacological molecular networks. Active components of S. miltiorrhiza depside salt and aspirin’s related genes were identified via the STITCH4.0 and GeneCards Database. A text search engine (Agilent Literature Search 2.71 and MCODE software were applied to construct network and divide modules, respectively. Finally, 32, 2, and 28 overlapping genes, modules, and pathways were identified between active components of S. miltiorrhiza depside salt and aspirin. A multidimensional framework of drug network showed that two networks reflected commonly in human aortic endothelial cells and atherosclerosis process. Aspirin plays a more important role in metabolism, such as the well-known AA metabolism pathway and other lipid or carbohydrate metabolism pathways. S. miltiorrhiza depside salt still plays a regulatory role in type II diabetes mellitus, insulin resistance, and adipocytokine signaling pathway. Therefore, this study suggests that aspirin combined with S. miltiorrhiza depside salt may be more efficient in treatment of CHD patients, especially those with diabetes mellitus or hyperlipidemia. Further clinical trials to confirm this hypothesis are still needed.

Evaluation of the neutron self-interrogation approach for assay of plutonium in high materials

International Nuclear Information System (INIS)

Russo, P.A.; Menlove, H.O.; Fife, K.W.; West, M.H.

1987-01-01

The pyrochemical scrap recovery processes, designed to extract impurities from plutonium metal and compounds, generate a variety of plutonium-laden residues consisting of high (α,n) matrices of varying chemical composition, and often containing grams to tens of grams of americium. For such materials, multiplication corrections based on real neutron coincidence count rate, R, and total neutron count rate, T, measurements cannot be applied because of the large, unknown, and variable (α,n) component in the total neutron emission rate. A study of the prototype self-interrogation assay method is in progress at the Los Alamos plutonium facility. In the self-interrogation approach, the assay signature R(IF)/T is a function of effective fissile plutonium content, where R(IF) is the induced fission component of the measured reals rate, and T is the measured, (α,n)-dominated totals rate. The present study includes a calibration effort using standards consisting of mixtures of PuO 2 and PuF 4 in a salt-strip matrix. The neutron measurements of the standards and the process materials have been performed at the Los Alamos Plutonium Facility. The precision and accuracy of the self-interrogation method applied to pyrochemical residues is examined in this study

Corrosion processes of alloyed steels in salt solutions

Energy Technology Data Exchange (ETDEWEB)

Kienzler, Bernhard [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Institut fuer Nukleare Entsorgung

2018-02-15

A summary is given of the corrosion experiments with alloyed Cr-Ni steels in salt solutions performed at Research Centre Karlsruhe (today KIT), Institute for Nuclear Waste Disposal (INE) in the period between 1980 and 2004. Alloyed steels show significantly lower general corrosion in comparison to carbon steels. However, especially in salt brines the protective Cr oxide layers on the surfaces of these steels are disturbed and localized corrosion takes place. Data on general corrosion rates, and findings of pitting, crevice and stress corrosion cracking are presented.

Alternatives for definse waste-salt disposal

International Nuclear Information System (INIS)

Benjamin, R.W.; McDonell, W.R.

1983-01-01

Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

[Monitoring of water and salt transport in silt and sandy soil during the leaching process].

Science.gov (United States)

Fu, Teng-Fei; Jia, Yong-Gang; Guo, Lei; Liu, Xiao-Lei

2012-11-01

Water and salt transport in soil and its mechanism is the key point of the saline soil research. The dynamic rule of water and transport in soil during the leaching process is the theoretical basis of formation, flush, drainage and improvement of saline soil. In this study, a vertical infiltration experiment was conducted to monitor the variation in the resistivity of silt and sandy soil during the leaching process by the self-designed automatic monitoring device. The experimental results showed that the peaks in the resistivity of the two soils went down and faded away in the course of leaching. It took about 30 minutes for sandy soil to reach the water-salt balance, whereas the silt took about 70 minutes. With the increasing leaching times, the desalination depth remained basically the same, being 35 cm for sandy soil and 10 cm for the silt from the top to bottom of soil column. Therefore, 3 and 7 leaching processes were required respectively for the complete desalination of the soil column. The temporal and spatial resolution of this monitoring device can be adjusted according to the practical demand. This device can not only achieve the remote, in situ and dynamic monitoring data of water and salt transport, but also provide an effective method in monitoring, assessment and early warning of salinization.

High temperature processes: from system reactivity to sensitive technology development and optimisation

International Nuclear Information System (INIS)

Lemont, F.

2007-01-01

The author gives a detailed overview of his scientific and research activities since 1997 in the fields of pyrochemical processes used to extract radio-elements from fission products produced by nuclear fuel reprocessing, and of the incineration of organic wastes contaminated by alpha emitters. He more precisely presents his works on the incineration of radioactive organic wastes by means of the IRIS process. Then, he discusses the actinide-lanthanide separation by a pyrometallurgical process. He also reports studies on the thermo-chemical cycles in the production of hydrogen

Effects of an Advocacy Trial on Food Industry Salt Reduction Efforts-An Interim Process Evaluation.

Science.gov (United States)

Trevena, Helen; Petersen, Kristina; Thow, Anne Marie; Dunford, Elizabeth K; Wu, Jason H Y; Neal, Bruce

2017-10-17

The decisions made by food companies are a potent factor shaping the nutritional quality of the food supply. A number of non-governmental organizations (NGOs) advocate for corporate action to reduce salt levels in foods, but few data define the effectiveness of advocacy. This present report describes the process evaluation of an advocacy intervention delivered by one Australian NGO directly to food companies to reduce the salt content of processed foods. Food companies were randomly assigned to intervention ( n = 22) or control ( n = 23) groups. Intervention group companies were exposed to pre-planned and opportunistic communications, and control companies to background activities. Seven pre-defined interim outcome measures provided an indication of the effect of the intervention and were assessed using intention-to-treat analysis. These were supplemented by qualitative data from nine semi-structured interviews. The mean number of public communications supporting healthy food made by intervention companies was 1.5 versus 1.8 for control companies ( p = 0.63). Other outcomes, including the mean number of news articles, comments and reports (1.2 vs. 1.4; p = 0.72), a published nutrition policy (23% vs. 44%; p = 0.21), public commitment to the Australian government's Food and Health Dialogue (FHD) (41% vs. 61%; p = 0.24), evidence of a salt reduction plan (23% vs. 30%; p = 0.56), and mean number of communications with the NGO (15 vs. 11; p = 0.28) were also not significantly different. Qualitative data indicated the advocacy trial had little effect. The absence of detectable effects of the advocacy intervention on the interim markers indicates there may be no impact of the NGO advocacy trial on the primary outcome of salt reduction in processed foods.

Geochemical processes in marine salt deposits: Their significance and their implications in connection with disposal of radioactive waste within salt domes

Energy Technology Data Exchange (ETDEWEB)

Herrmann, A G [Goettingen Univ. (Germany, F.R.). Geochemisches Inst.

1980-01-01

Attempts to effect permanent disposal of radioactive wastes in marine evaporites should do nothing to disturb, either in the short or the long term, the present relative stability of such bodies of rock. It is necessary to take account of all of the geochemical and physico-chemical reactions known to have been involved in the processes which formed the evaporites before proceeding to an acceptable strategy for disposal of radionucleides. These processes can be represented as three kinds of metamorphism: 1. solution metamorphism, 2. thermal metamorphism, 3. dynamic metamorphism. In all of the evaporite occurrences in Germany such processes have been influential in altering, on occasion significantly, the primary mineralogical composition and have also promoted a considerable degree of transposition of material. Given similar geochemical and physico-chemical premises, these metamorphic processes could become effective now or in the future. It is therefore necessary to discuss the following criteria when examining salt domes as permanent repositories of highly radioactive substances: (1) Temperatures <= 90/sup 0/ +- 10/sup 0/C at the contact between waste containers and rock salt; (2) Temperatures <= 75/sup 0/C within zones of carnallite rocks; (3) Immobilisation of high-level waste in crystalline forms whenever possible; (4) Systems of additional safety barriers around the waste containers or the unreprocessed spent fuel elements. The geochemical and physical effectiveness of the barriers within an evaporite environment must be guaranteed. For example: Ni-Ti-alloys, corundum, ceramic, anhydrite.

Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

International Nuclear Information System (INIS)

Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

2008-01-01

An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

Chlorination of UO2, PuO2, and rare-earth oxides using ZrCl4

International Nuclear Information System (INIS)

Sakamura, Yoshiharu; Inoue, Tadashi; Iwai, Takashi; Moriyama, Hirotake

2001-01-01

A new chlorination method using ZrCl 4 , which has a high reactivity with oxygen, has been investigated for more efficient oxide treatment. After actinide oxides are chlorinated and dissolved in a molten salt bath, actinide metals can be selectively collected using the electrorefining process. This process is well suited for pyrochemical reprocessing of metallic fuels. In LiCl-KCI eutectic melts, rare-earth oxides (Y 2 O 3 , La 2 O 3 , CeO 2 , and Nd 2 O 3 ) and actinide oxides (UO 2 and PuO 2 ) were chlorinated by adding ZrCl 4 . As a result, rare-earth and actinide elements were dissolved into the salt as trivalent ions and ZrO 2 was precipitated. When an excess of ZrCI 4 was added, oxides in powder form were completely chlorinated in five hours. It was demonstrated that the ZrCI 4 chlorination method, free from corrosive gas such as chlorine, was very simple and useful. (author)

Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

Energy Technology Data Exchange (ETDEWEB)

Peters, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-09-07

The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

Assessment by X-ray diffraction the process of bentonite organophilization using a different quaternary ammonium salts

International Nuclear Information System (INIS)

Silva, L.A. da; Rosario, J.A. do; Lima, R.B.; Milioli, C.C.; Gusatti, M.; Linhares, R.H.; Kuhnen, N.C.; Riella, H.G.

2010-01-01

The process was conducted in an organophilization Bentonite originated from the Company of Industrial Minerals of Mozambique Ltd. (Mimoc). The transformation of bentonite organophilic clay were performed in laboratory procedures that aim to mechanochemical exchange of Na + and Ca 2+ from the interlayer space of clay minerals by cations of quaternary ammonium salts. In this study we used two types of salts, which are: the cetyl trimethyl ammonium chloride and alkyl dimethyl benzyl ammonium chloride at different concentrations (30, 50, 80, 100 meq/100 g clay). The natural bentonite and organophilic clay samples were characterized by X-ray diffraction (XRD) to obtain the mineralogical constituents and analysis phases of the increase in interlayer distance confirming the incorporation of quaternary ammonium salts in the structure of clays. (author)

Research and development of pyrochemical technologies for ADTT systems in NRI Rez plc

International Nuclear Information System (INIS)

Uhlir, J.

1999-01-01

The activities within the Czech national Partitioning and Transmutation program are oriented mainly toward the following areas: (i) ADS flowsheeting working out a proposal for a suitable technological flowsheet for treating spent fuel into a form fitted for transmutation reactor, including separation procedures before transmutation (Front-end) and separation processes after passage of fuel through the transmutor (Back-end). (ii) Technological research into the uranium component separation from spent fuel by the Fluoride Volatility Method semi-pilot-plant technological research based on the application of experience from spent fuel fluoride reprocessing. (iii) Laboratory research into electro-separation methods for the separation of fission products, plutonium and minor actinides from the fluoride melt. Selection of a suitable matrix composition of fluoride melts based on studies of their physicochemical properties and design of laboratory electro-separation facility. (iv) Development of materials and equipment for molten fluoride salts based ADS technologies, i.e. construction materials development, research into the corrosion resistance and development of equipment for molten fluoride salts technologies. The NRI Rez plc cooperates in this area with the SKODA Nuclear Machinery company

Potential effect of salt reduction in processed foods on health

NARCIS (Netherlands)

Hendriksen, M.A.H.; Hoogenveen, R.T.; Hoekstra, J.; Geleijnse, J.M.; Boshuizen, H.C.; Raaij, van J.M.A.

2014-01-01

Background: Excessive salt intake has been associated with hypertension and increased cardiovascular disease morbidity and mortality. Reducing salt intake is considered an important public health strategy in the Netherlands. Objective: The objective was to evaluate the health benefits of

Process for improving the energy density of feedstocks using formate salts

Energy Technology Data Exchange (ETDEWEB)

Wheeler, Marshall Clayton; van Heiningen, Adriaan R.P.; Case, Paige A.

2015-09-01

Methods of forming liquid hydrocarbons through thermal deoxygenation of cellulosic compounds are disclosed. Aspects cover methods including the steps of mixing a levulinic acid salt-containing feedstock with a formic acid salt, exposing the mixture to a high temperature condition to form hydrocarbon vapor, and condensing the hydrocarbon vapor to form liquid hydrocarbons, where both the formic acid salt and the levulinic acid salt-containing feedstock decompose at the high temperature condition and wherein one or more of the mixing, exposing, and condensing steps is carried out a pressure between about vacuum and about 10 bar.

Numerical analysis of impurity separation from waste salt by investigating the change of concentration at the interface during zone refining process

Science.gov (United States)

Choi, Ho-Gil; Shim, Moonsoo; Lee, Jong-Hyeon; Yi, Kyung-Woo

2017-09-01

The waste salt treatment process is required for the reuse of purified salts, and for the disposal of the fission products contained in waste salt during pyroprocessing. As an alternative to existing fission product separation methods, the horizontal zone refining process is used in this study for the purification of waste salt. In order to evaluate the purification ability of the process, three-dimensional simulation is conducted, considering heat transfer, melt flow, and mass transfer. Impurity distributions and decontamination factors are calculated as a function of the heater traverse rate, by applying a subroutine and the equilibrium segregation coefficient derived from the effective segregation coefficients. For multipass cases, 1d solutions and the effective segregation coefficient obtained from three-dimensional simulation are used. In the present study, the topic is not dealing with crystal growth, but the numerical technique used is nearly the same since the zone refining technique was just introduced in the treatment of waste salt from nuclear power industry because of its merit of simplicity and refining ability. So this study can show a new application of single crystal growth techniques to other fields, by taking advantage of the zone refining multipass possibility. The final goal is to achieve the same high degree of decontamination in the waste salt as in zone freezing (or reverse Bridgman) method.

Influence of the Chemical Interactions on the Removal Rate of Different Salts in Electrokinetic Desalination Processes

DEFF Research Database (Denmark)

Paz-Garcia, Juan Manuel; Johannesson, Björn; Ottosen, Lisbeth M.

2011-01-01

Electrokinetic desalination techniques have been successfully applied for the prevention of salt-induced deterioration problems of masonry and other construction materials. A mathematical model for electrochemical desalination treatments is described, based on the Poisson-Nernst-Planck system...... of equations and accounting for the chemical interactions between the species in the pore solution and the solid matrix. Due to their high abundance in the natural environment, chlorides, nitrates and sulfates are considered the main ions responsible to the salt decay processes in buildings materials...

Enhancing and accelarating flavour formation by salt-tolerant yeasts in Japanese soy-sauce processes

NARCIS (Netherlands)

Sluis, van der C.; Tramper, J.; Wijffels, R.H.

2001-01-01

In soy-sauce processes salt-tolerant yeasts are very important for the flavour formation. This flavour formation is, however, slow and poorly understood. In the last decades, a concerted research effort has increased the understanding and resulted in the derivation of mutants with an enhanced

Salt tectonics in Santos Basin, Brazil

Energy Technology Data Exchange (ETDEWEB)

Quirk, David G.; Nielsen, Malene; Raven, Madeleine [Maersk Oil and Gas, Copenhagen (Denmark); Menezes, Paulo [Maersk Oil and Gas, Rio de Janeiro, RJ (Brazil)

2008-07-01

From Albian to end Cretaceous times, the inboard part of the Santos Basin in Brazil was affected by extension as salt flowed basinwards under the effect of gravity. Salt rollers, flip-flop salt diapirs and the famous Albian Gap were all formed by this process. Outboard of these extensional structures, contraction was taken up in a wide zone of thickened salt where salt collected. The overburden was carried on top of the salt as it flowed down-dip, with up to 40 km of translation recorded in Albian strata. (author)

Chemistry and technology of Molten Salt Reactors - history and perspectives

International Nuclear Information System (INIS)

Uhlir, Jan

2007-01-01

Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

Process and apparatus for extraction of gases produced during operation of a fused-salt nuclear reactor

International Nuclear Information System (INIS)

Blum, J.; Marie, J.

1976-01-01

The present invention relates to the field of fused-salt nuclear reactors and its object is the extraction of the gases produced in the course of operation of these reactors. The process according to the invention consists in placing into position a piece of material permeable for gases and impermeable for the used fused salts, for instance, a piece of graphite, in such a way that part of the surface of this piece is in contact with the circuit of the radioactive salts and another part connected to a gas suction device. The piece could also be scavenged in its mass by a flow of inert gas. Application is contemplated in reactors using a mixture of lithium fluoride, beryllium fluoride, and uranium and/or thorium fluoride. 10 claims, 2 drawing figures

Phase Equilibrium Studies of Savannah River Tanks and Feed Streams for the Salt Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

Weber, C.F.

2001-06-19

A chemical equilibrium model is developed and used to evaluate supersaturation of tanks and proposed feed streams to the Salt Waste Processing Facility. The model uses Pitzer's model for activity coefficients and is validated by comparison with a variety of thermodynamic data. The model assesses the supersaturation of 13 tanks at the Savannah River Site (SRS), indicating that small amounts of gibbsite and or aluminosilicate may form. The model is also used to evaluate proposed feed streams to the Salt Waste Processing Facility for 13 years of operation. Results indicate that dilutions using 3-4 M NaOH (about 0.3-0.4 L caustic per kg feed solution) should avoid precipitation and reduce the Na{sup +} ion concentration to 5.6 M.

The source term and waste optimization of molten salt reactors with processing

International Nuclear Information System (INIS)

Gat, U.; Dodds, H.L.

1993-01-01

The source term of a molten salt reactor (MSR) with fuel processing is reduced by the ratio of processing time to refueling time as compared to solid fuel reactors. The reduction, which can be one to two orders of magnitude, is due to removal of the long-lived fission products. The waste from MSRs can be optimized with respect to its chemical composition, concentration, mixture, shape, and size. The actinides and long-lived isotopes can be separated out and returned to the reactor for transmutation. These features make MSRs more acceptable and simpler in operation and handling

Residual Salt Separation from the Metal Products Reduced in a LiCl-Li2O Molten Salt

International Nuclear Information System (INIS)

Hur, Jin Mok; Hong, Sun Seok; Kang, Dae Seung; Jeong, Meong Soo; Seo, Chung Seok

2006-02-01

The electrochemical reduction of spent nuclear fuel in a LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for the active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel components which might be metallized by the electrochemical reduction process. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of a LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950 .deg. C were measured as follows: LiCl-8 wt% Li 2 O > LiCl > LiCl-8 wt% SrCl 2 > SrCl 2

Potential Process for the Decontamination of Pyro-electrometallurgical LiCl-KCl Eutectic Salt Electrolyte

International Nuclear Information System (INIS)

Griffith, Christopher S.; Sizgek, Erden; Sizgek, Devlet; Luca, Vittorio

2008-01-01

Presented here is a potential option with experimental validation for the decontamination of LiCl-KCl eutectic salt electrolyte from a pyro-electrometallurgical process by employing already developed inorganic ion exchange materials. Adsorbent materials considered include titano-silicates and molybdo- and tungstophosphates for Cs extraction, Si-doped antimony pyrochlore for Sr extraction and hexagonal tungsten bronzes for lanthanide (LN) and minor actinide (MA) polishing. Encouraging results from recent investigations on the removal of target elements (Cs, Sr and LN) from aqueous solutions containing varying concentrations of alkali and alkali metal contaminants which would be akin to a solution formed from the dissolution of spent LiCl-KCl eutectic salt electrolyte are presented. Further investigations have also shown that the saturated adsorbents can be treated at relatively low temperatures to afford potential waste forms for the adsorbed elements. Efficient evaporation and drying of a solution of dissolved LiCl-KCl eutectic salt electrolyte (50 L, 5 L.h -1 ) has been demonstrated using a Microwave-Heated Mechanical Fluidized Bed (MWMFB) apparatus. (authors)

Potential Process for the Decontamination of Pyro-electrometallurgical LiCl-KCl Eutectic Salt Electrolyte

Energy Technology Data Exchange (ETDEWEB)

Griffith, Christopher S.; Sizgek, Erden; Sizgek, Devlet; Luca, Vittorio [Australian Nuclear Science and Technology Organisation (ANSTO), Institute of Materials Engineering, New Illawarra Road, Lucas Heights, New South Wales, 2234 (Australia)

2008-07-01

Presented here is a potential option with experimental validation for the decontamination of LiCl-KCl eutectic salt electrolyte from a pyro-electrometallurgical process by employing already developed inorganic ion exchange materials. Adsorbent materials considered include titano-silicates and molybdo- and tungstophosphates for Cs extraction, Si-doped antimony pyrochlore for Sr extraction and hexagonal tungsten bronzes for lanthanide (LN) and minor actinide (MA) polishing. Encouraging results from recent investigations on the removal of target elements (Cs, Sr and LN) from aqueous solutions containing varying concentrations of alkali and alkali metal contaminants which would be akin to a solution formed from the dissolution of spent LiCl-KCl eutectic salt electrolyte are presented. Further investigations have also shown that the saturated adsorbents can be treated at relatively low temperatures to afford potential waste forms for the adsorbed elements. Efficient evaporation and drying of a solution of dissolved LiCl-KCl eutectic salt electrolyte (50 L, 5 L.h{sup -1}) has been demonstrated using a Microwave-Heated Mechanical Fluidized Bed (MWMFB) apparatus. (authors)

Residual salts separation from metal reduced electrolytically in a LiCl-Li2O molten salt

International Nuclear Information System (INIS)

Hur, Jin Mok; Oh, Seung Chul; Hong, Sun Seok; Seo, Chung Seok; Park, Seong Won

2005-01-01

The PWR spent oxide fuel can be reduced electrolytically in a hot molten salt for the conditioning and the preparation of a metallic fuel. Then the metal product is smelted into an ingot to be treated in the post process. Incidentally, the residual salt which originated from the molten salt and spent fuel elements should be separated from the metal product during the smelting. In this work, we constructed a surrogate material system to simulate the salt separation from the reduced spent fuel and studied the vaporization behaviors of the salts

Production of carboxylic acid and salt co-products

Science.gov (United States)

Hanchar, Robert J.; Kleff, Susanne; Guettler, Michael V.

2014-09-09

This invention provide processes for producing carboxylic acid product, along with useful salts. The carboxylic acid product that is produced according to this invention is preferably a C.sub.2-C.sub.12 carboxylic acid. Among the salts produced in the process of the invention are ammonium salts.

Fracture and Healing of Rock Salt Related to Salt Caverns

International Nuclear Information System (INIS)

Chan, K.S.; Fossum, A.F.; Munson, D.E.

1999-01-01

In recent years, serious investigations of potential extension of the useful life of older caverns or of the use of abandoned caverns for waste disposal have been of interest to the technical community. All of the potential applications depend upon understanding the reamer in which older caverns and sealing systems can fail. Such an understanding will require a more detailed knowledge of the fracture of salt than has been necessary to date. Fortunately, the knowledge of the fracture and healing of salt has made significant advances in the last decade, and is in a position to yield meaningful insights to older cavern behavior. In particular, micromechanical mechanisms of fracture and the concept of a fracture mechanism map have been essential guides, as has the utilization of continuum damage mechanics. The Multimechanism Deformation Coupled Fracture (MDCF) model, which is summarized extensively in this work was developed specifically to treat both the creep and fracture of salt, and was later extended to incorporate the fracture healing process known to occur in rock salt. Fracture in salt is based on the formation and evolution of microfractures, which may take the form of wing tip cracks, either in the body or the boundary of the grain. This type of crack deforms under shear to produce a strain, and furthermore, the opening of the wing cracks produce volume strain or dilatancy. In the presence of a confining pressure, microcrack formation may be suppressed, as is often the case for triaxial compression tests or natural underground stress situations. However, if the confining pressure is insufficient to suppress fracture, then the fractures will evolve with time to give the characteristic tertiary creep response. Two first order kinetics processes, closure of cracks and healing of cracks, control the healing process. Significantly, volume strain produced by microfractures may lead to changes in the permeability of the salt, which can become a major concern in

Projected Salt Waste Production from a Commercial Pyroprocessing Facility

Directory of Open Access Journals (Sweden)

Michael F. Simpson

2013-01-01

Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

Science.gov (United States)

Merk, B; Litskevich, D; Gregg, R; Mount, A R

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

System and process for production of magnesium metal and magnesium hydride from magnesium-containing salts and brines

Science.gov (United States)

McGrail, Peter B.; Nune, Satish K.; Motkuri, Radha K.; Glezakou, Vassiliki-Alexandra; Koech, Phillip K.; Adint, Tyler T.; Fifield, Leonard S.; Fernandez, Carlos A.; Liu, Jian

2016-11-22

A system and process are disclosed for production of consolidated magnesium metal products and alloys with selected densities from magnesium-containing salts and feedstocks. The system and process employ a dialkyl magnesium compound that decomposes to produce the Mg metal product. Energy requirements and production costs are lower than for conventional processing.

Artisanal salt production in Aveiro/Portugal - an ecofriendly process.

Science.gov (United States)

Rodrigues, Carolina M; Bio, Ana; Amat, Francisco; Vieira, Natividade

2011-11-04

Solar salinas are man-made systems exploited for the extraction of salt, by solar and wind evaporation of seawater. Salt production achieved by traditional methods is associated with landscapes and environmental and patrimonial values generated throughout history. Since the mid-twentieth century, this activity has been facing a marked decline in Portugal, with most salinas either abandoned or subjected to destruction, making it necessary to find a strategy to reverse this trend.It is, however, possible to generate revenue from salinas at several levels, not merely in terms of good quality salt production, but also by obtaining other products that can be commercialized, or by exploring their potential for tourism, and as research facilities, among others. Furthermore, with an adequate management, biodiversity can be restored to abandoned salinas, which constitute important feeding and breeding grounds for resident and migratory aquatic birds, many of which are protected by European Community Directives.The aims of this manuscript are to present a brief overview on the current state of sea salt exploitation in Portugal and to stress the importance of recovering these salinas for the conservation of this particular environment, for the regional economy, the scientific community and the general public. The Aveiro salina complex is presented in detail, to exemplify salina structure and functioning, as well as current problems and potential solutions for artisanal salinas.

Tanks Focus Area Alternative Salt Processing Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, Harry D.

2000-05-15

In March 2000, DOE-Headquarters (HQ) requested the Tanks Focus Area (TFA)to assume management responsibility for the Salt Processing Project technology development program at Savannah River Site. The TFA was requested to conduct several activities, including review and revision of the technology development roadmaps, development of down-selection criteria, and preparation of a comprehensive Research and Development (R&D) Program Plan for three candidate cesium removal technologies, as well as the Alpha and strontium removal processes that must also be carried out. The three cesium removal candidate technologies are Crystalline Silicotitanate (CST) Non-Elutable Ion Exchange, Caustic Side Solvent Extraction (CSSX), and Small Tank Tetraphenylborate Precipitation (STTP). This plan describes the technology development needs for each process that must be satisfied in order to reach a down-selection decision, as well as continuing technology development required to support conceptual design activities.

Tanks Focus Area Alternative Salt Processing Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, Harry D.

2000-11-30

In March 2000, DOE-Headquarters (HQ) requested the Tanks Focus Area (TFA) to assume management responsibility for the Salt Processing Project technology development program at Savannah River Site. The TFA was requested to conduct several activities, including review and revision of the technology development roadmaps, development of down-selection criteria, and preparation of a comprehensive Research and Development (R&D) Program Plan for three candidate cesium removal technologies, as well as the Alpha and strontium removal processes that must also be carried out. The three cesium removal candidate technologies are Crystalline Silicotitanate (CST) Non-Elutable Ion Exchange, Caustic Side Solvent Extraction (CSSX), and Small Tank Tetraphenylborate Precipitation (STTP). This plan describes the technology development needs for each process that must be satisfied in order to reach a down-selection decision, as well as continuing technology development required to support conceptual design activities.

Individual aerosol particles in and below clouds along a Mt. Fuji slope: Modification of sea-salt-containing particles by in-cloud processing

Science.gov (United States)

Ueda, S.; Hirose, Y.; Miura, K.; Okochi, H.

2014-02-01

Sizes and compositions of atmospheric aerosol particles can be altered by in-cloud processing by absorption/adsorption of gaseous and particulate materials and drying of aerosol particles that were formerly activated as cloud condensation nuclei. To elucidate differences of aerosol particles before and after in-cloud processing, aerosols were observed along a slope of Mt. Fuji, Japan (3776 m a.s.l.) during the summer in 2011 and 2012 using a portable laser particle counter (LPC) and an aerosol sampler. Aerosol samples for analyses of elemental compositions were obtained using a cascade impactor at top-of-cloud, in-cloud, and below-cloud altitudes. To investigate composition changes via in-cloud processing, individual particles (0.5-2 μm diameter) of samples from five cases (days) collected at different altitudes under similar backward air mass trajectory conditions were analyzed using a transmission electron microscope (TEM) equipped with an energy dispersive X-ray analyzer. For most cases (four cases), most particles at all altitudes mainly comprised sea salts: mainly Na with some S and/or Cl. Of those, in two cases, sea-salt-containing particles with Cl were found in below-cloud samples, although sea-salt-containing particles in top-of-cloud samples did not contain Cl. This result suggests that Cl in the sea salt was displaced by other cloud components. In the other two cases, sea-salt-containing particles on samples at all altitudes were without Cl. However, molar ratios of S to Na (S/Na) of the sea-salt-containing particles of top-of-cloud samples were higher than those of below-cloud samples, suggesting that sulfuric acid or sulfate was added to sea-salt-containing particles after complete displacement of Cl by absorption of SO2 or coagulation with sulfate. The additional volume of sulfuric acid in clouds for the two cases was estimated using the observed S/Na values of sea-salt-containing particles. The estimation revealed that size changes by in

Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

International Nuclear Information System (INIS)

Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

2012-01-01

The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

“Use salt and foods high in salt sparingly”: A food-based dietary ...

African Journals Online (AJOL)

Legislating the levels of salt in processed food is only one part of this national strategy. All health professionals and educators should also provide appropriate nutritional recommendations that will educate, motivate and enable consumers to change their nutritional behaviour to reduce salt intake to less than 5 g per day, ...

PRODUCTION OF INDUSTRY SALT WITH SEDIMENTATION – MICROFILTRATION PROCESS: OPTIMAZATION OF TEMPERATURE AND CONCENTRATION BY USING SURFACE RESPONSE METHODOLOGY

Directory of Open Access Journals (Sweden)

Widayat Widayat

2012-02-01

Full Text Available The salt of sodium chloride commonly used consumption in house , so as a raw material in industry. Thequality of salt depends on sodium chloride concentration. The objective of this research is obtained ofoptimum condition in production of salt industry by using sedimentation and microfiltration process. Theoptimization used surface response methodology and analysis by Statistica 6 software. The responseperceived is NaCl concentration in product. The experiments do by mixing stearic acid with NaOH solutionto product stearic sodium. Then, the solution mixed with sea water, so the white solid will be emerge, thereare stearic calcium and stearic magnesium. And so filtrate evaporated until to obtain salt. TheMathematical model for reduction of Ca2+ and Mg2+ are1 222 221 1 Y = 93,3185 + 1,0967 X + 0,1909 X +1,0682 X - 0,2333 X - 0,3376 X X , with maximum conversion is94,46% at temperature 82,42oC and stearic sodium concentration 14,16%(v/v. The maximum of NaClconcentration is 96,19% at temperature 81,54oC and stearic sodium concetration 13,11 %(v/v. Themathematical model for NaCl production is1 222 221 1 Y = 92,7596 − 0,3443 X − 3,3706 X + 2,9553 X - 0,9562 X - 1,9272 X X . The results of NaCl not yetfulfilled with SNI industry salt. The NaCl concetration in SNI is 98,5%. So, this process is nt aplicable forproductiob salt industry in Indonesia.

Salt ingestion caves.

Directory of Open Access Journals (Sweden)

Lundquist Charles A.

2006-01-01

Full Text Available Large vertebrate herbivores, when they find a salt-bearing layer of rock, say in a cliff face, can produce sizable voids where, overgenerations, they have removed and consumed salty rock. The cavities formed by this natural animal process constitute a uniqueclass of caves that can be called salt ingestion caves. Several examples of such caves are described in various publications. Anexample in Mississippi U.S.A., Rock House Cave, was visited by the authors in 2000. It seems to have been formed by deer orbison. Perhaps the most spectacular example is Kitum Cave in Kenya. This cave has been excavated to a length over 100 metersby elephants. An ancient example is La Cueva del Milodon in Chile, which is reported to have been excavated by the now extinctmilodon, a giant ground sloth. Still other possible examples can be cited. This class of caves deserves a careful definition. First, thecavity in rock should meet the size and other conventions of the locally accepted definition of a cave. Of course this requirement differsin detail from country to country, particularly in the matter of size. The intent is to respect the local conventions. The characteristicthat human entry is possible is judged to be a crucial property of any recognized cave definition. Second, the cavity should besignificantly the result of vertebrate animal consumption of salt-bearing rock. The defining process is that rock removed to form thecave is carried away in the digestive track of an animal. While sodium salts are expected to be the norm, other salts for which thereis animal hunger are acceptable. Also some other speleogenesis process, such as solution, should not be excluded as long as it issecondary in formation of a cave in question.

Separation of adhered salt from uranium deposits generated in electro-refiner

International Nuclear Information System (INIS)

Kwon, S.W.; Park, K.M.; Lee, H.S.; Kim, J.G.; Ahn, H.G.

2011-01-01

It is important to increase a throughput of the salt removal process from uranium deposits which is generated on the solid cathode of electro-refiner in pyroprocess. In this study, it was proposed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl-KCl eutectic salt from the uranium deposits. The feasibility of liquid salt separation was examined by salt separation experiments on a stainless steel sieve. It was found that the amount of salt to be distilled could be reduced by the liquid salt separation prior to the salt distillation. The residual salt remained in the deposits after the liquid salt separation was successfully removed further by the vacuum distillation. It was concluded that the combination of a liquid salt separation and a vacuum distillation is an effective route for the achievement of a high throughput performance in the salt separation process. (author)

Completion report for the UMTRA project Vitro processing site, Salt Lake City, Utah

International Nuclear Information System (INIS)

1996-08-01

This completion report provides evidence that the final Salt Lake City, Utah, processing site property conditions are in accordance with the approval design and that all U.S. Environmental Protection Agency (EPA) standards have been satisfied. Included as appendixes to support the stated conclusions are the record drawings; a summary of grid test results; contract specifications and construction drawing and the EPA standards; the audit, inspection, and surveillance summary; the permit information; and project photographs. The principal objectives of remedial action at Salt Lake City were to remove the tailings from the former processing site, render the site free of contamination to EPA standards, and restore the site to the final design grade elevations. The final remedial action plan, which is approved by the U.S. Department of Energy and concurred upon by the U.S. Nuclear Regulator Commission and the state of Utah, contains the conceptual design used to develop the final approved design. During remedial action construction operations, conditions were encountered that required design features that differed form the conceptual design. These conditions and the associated design changes are noted in the record drawings. All remedial action activities were completed in conformance with the specifications and drawings. In the opinion of the state of Utah, the record drawings accurately reflect existing property conditions at the processing site

Salt fog corrosion behavior in a powder-processed icosahedral-phase-strengthened aluminum alloy

International Nuclear Information System (INIS)

Watson, T.J.; Gordillo, M.A.; Ernst, A.T.; Bedard, B.A.; Aindow, M.

2017-01-01

Highlights: • Pitting corrosion resistance has been evaluated for an Al-Cr-Mn-Co-Zr alloy. • Pit densities and depths are far lower than for other high-strength Al alloys. • Corrosion proceeds by selective oxidation of the Al matrix around the other phases. - Abstract: The pitting corrosion resistance has been evaluated for a powder-processed Al-Cr-Mn-Co-Zr alloy which contains ≈35% by volume of an icosahedral quasi-crystalline phase and a little Al 9 Co 2 in an Al matrix. ASTM standard salt fog exposure tests show that the alloy exhibits far lower corrosion pit densities and depths than commercial high-strength aerospace Al alloys under the same conditions. Electron microscopy data show that the salt fog exposure leads to the selective oxidation of the face-centered cubic Al matrix around the other phases, and to the development of a porous outer oxide scale.

The Results of HLW Processing Using Zirconium Salt of Dibutyl phosphoric Acid in Hot Cell

Energy Technology Data Exchange (ETDEWEB)

Fedorov, Yu.S.; Zilberman, B.Ya.; Shmidt, O.V. [Khlopin Radium Institute, 2nd Murinsky Ave., 28, Saint-Petersburg, 194021 (Russian Federation)

2008-07-01

Zirconium salt of dibutyl phosphoric acid (ZS HDBP), is an effective solvent for liquid HLW and ILW (high and intermediate level wastes) processing with radionuclide partitioning into different groups for further immobilization according to radiotoxicity. The rig trials in mixer-settles in hot cells were carried out using 30 L of real HLW containing transplutonium (TPE), rare earths (RE), Sr and Cs in 2 mol/L HNO{sub 3}, characterized by total specific activity 520 MBk/L. The recovery factor for TPE and RE was as high as 10{sup 4}, but only 10 for Sr. Purification factor of TPE and RE from Cs and Sr was 10{sup 4}, and that of Sr from TPE and Cs was 10{sup 3}. Almost all Cs was localized in the second cycle raffinate. So Zr salt of HDBP can be used in HLW processing with radionuclide partitioning with respect to the categories of radiotoxicity. (authors)

Brine flow in heated geologic salt.

Energy Technology Data Exchange (ETDEWEB)

Kuhlman, Kristopher L.; Malama, Bwalya

2013-03-01

This report is a summary of the physical processes, primary governing equations, solution approaches, and historic testing related to brine migration in geologic salt. Although most information presented in this report is not new, we synthesize a large amount of material scattered across dozens of laboratory reports, journal papers, conference proceedings, and textbooks. We present a mathematical description of the governing brine flow mechanisms in geologic salt. We outline the general coupled thermal, multi-phase hydrologic, and mechanical processes. We derive these processes governing equations, which can be used to predict brine flow. These equations are valid under a wide variety of conditions applicable to radioactive waste disposal in rooms and boreholes excavated into geologic salt.

Uranium (III) precipitation in molten chloride by wet argon sparging

Energy Technology Data Exchange (ETDEWEB)

Vigier, Jean-François, E-mail: jean-francois.vigier@ec.europa.eu [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Laplace, Annabelle [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Renard, Catherine [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Miguirditchian, Manuel [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Abraham, Francis [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France)

2016-06-15

In the context of pyrochemical processes for nuclear fuel treatment, the precipitation of uranium (III) in molten salt LiCl-CaCl{sub 2} (30–70 mol%) at 705 °C is studied. First, this molten chloride is characterized with the determination of the water dissociation constant. With a value of 10{sup −4.0}, the salt has oxoacid properties. Then, the uranium (III) precipitation using wet argon sparging is studied. The salt is prepared using UCl{sub 3} precursor. At the end of the precipitation, the salt is totally free of solubilized uranium. The main part is converted into UO{sub 2} powder but some uranium is lost during the process due to the volatility of uranium chloride. The main impurity of the resulting powder is calcium. The consequences of oxidative and reductive conditions on precipitation are studied. Finally, coprecipitation of uranium (III) and neodymium (III) is studied, showing a higher sensitivity of uranium (III) than neodymium (III) to precipitation. - Highlights: • Precipitation of Uranium (III) is quantitative in molten salt LiCl-CaCl{sub 2} (30–70 mol%). • The salt is oxoacid with a water dissociation constant of 10{sup −4.0} at 705 °C. • Volatility of uranium chloride is strongly reduced in reductive conditions. • Coprecipitation of U(III) and Nd(III) leads to a consecutive precipitation of the two elements.

Residual Salt Separation from the Metal Products Reduced in a LiCl-Li{sub 2}O Molten Salt

Energy Technology Data Exchange (ETDEWEB)

Hur, Jin Mok; Hong, Sun Seok; Kang, Dae Seung; Jeong, Meong Soo; Seo, Chung Seok

2006-02-15

The electrochemical reduction of spent nuclear fuel in a LiCl-Li{sub 2}O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for the active tests. Fresh uranium metal prepared from the electrochemical reduction of U{sub 3}O{sub 8} powder was used as the surrogates of the spent nuclear fuel components which might be metallized by the electrochemical reduction process. LiCl, Li{sub 2}O, Y{sub 2}O{sub 3} and SrCl{sub 2} were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li{sub 2}O and LiCl-SrCl{sub 2} led to a melting point which was lower than that of a LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li{sub 2}O and LiCl-SrCl{sub 2} was achieved below temperatures which could make the uranium metal oxidation by Li{sub 2}O possible. The salt vaporization rates at 950 .deg. C were measured as follows: LiCl-8 wt% Li{sub 2}O > LiCl > LiCl-8 wt% SrCl{sub 2} > SrCl{sub 2}.

Modeling Episodic Ephemeral Brine Lake Evaporation and Salt Crystallization on the Bonneville Salt Flats, Utah

Science.gov (United States)

Liu, T.; Harman, C. J.; Kipnis, E. L.; Bowen, B. B.

2017-12-01

Public concern about apparent reductions in the areal extent of the Bonneville Salt Flat (BSF) and perceived changes in inundation frequency has motivated renewed interest in the hydrologic and geochemical behavior of this salt playa. In this study, we develop a numerical modeling framework to simulate the relationship between hydrometeorologic variability, brine evaporation and salt crystallization processes on BSF. The BSF, locates in Utah, is the remnant of paleo-lake Bonneville, and is capped by up to 1 meter of salt deposition over a 100 km2 area. The BSF has two distinct hydrologic periods each year: a winter wet periods with standing surface brine and the summer dry periods when the brine is evaporated, exposing the surface salt crust. We develop a lumped non-linear dynamical models coupling conservation expressions from water, dissolved salt and thermal energy to investigate the seasonal and diurnal behavior of brine during the transition from standing brine to exposed salt at BSF. The lumped dynamic models capture important nonlinear and kinetic effects introduced by the high ionic concentration of the brine, including the pronounced effect of the depressed water activity coefficient on evaporation. The salt crystallization and dissolution rate is modeled as a kinetic process linearly proportional to the degree of supersaturation of brine. The model generates predictions of the brine temperature and the solute and solvent masses controlled by diurnal net radiation input and aerodynamic forcing. Two distinct mechanisms emerge as potential controls on salt production and dissolution: (1) evapo-concentration and (2) changes in solubility related to changes in brine temperature. Although the evaporation of water is responsible for ultimate disappearance of the brine each season ,variation in solubility is found to be the dominant control on diurnal cycles of salt precipitation and dissolution in the BSF case. Most salt is crystallized during nighttime, but the

Salt briquette: the form of salt monopoly in madura, 1883-1911

Science.gov (United States)

Wisnu; Alrianingrum, S.; Artono; Liana, C.

2018-01-01

This study describes the history of the salt monopoly in Indonesia because it is associated with the issue of salt crisis lately, widely reported in various media. This study tried to find answers to the relationship between monopoly and crisis events through the study of history. Monopoly policy by the government of the colonial period is actually an industrial modernization effort, but it turned out another impact. Although the colonial government wanted to issue a policy that ends strengthens the position of the government in the industry, but ultimately backfire and disasters in the salt industry at the time. This article discusses only the focus of the salt monopoly in Madura as a selection of events, arguing the island as a center of salt in Indonesia. The method used in this study using a review of history. Therefore, their explanations using historical sources. Methodologically through the process of collecting historical sources, criticize these sources, synthesize and interpret the analysis in an array of historical writing. In conclusion, although the salt monopoly policy gives a great advantage to the colonial government, but the overall population of Madura remains in a poor state. It is evident that the Madurese to migrate Madurese to various areas outside the island of Madura, to fix the economy.

Submarine Salt Karst Terrains

Directory of Open Access Journals (Sweden)

Nico Augustin

2016-06-01

Full Text Available Karst terrains that develop in bodies of rock salt (taken as mainly of halite, NaCl are special not only for developing in one of the most soluble of all rocks, but also for developing in one of the weakest rocks. Salt is so weak that many surface-piercing salt diapirs extrude slow fountains of salt that that gravity spread downslope over deserts on land and over sea floors. Salt fountains in the deserts of Iran are usually so dry that they flow at only a few cm/yr but the few rain storms a decade so soak and weaken them that they surge at dm/day for a few days. We illustrate the only case where the rates at which different parts of one of the many tens of subaerial salt karst terrains in Iran flows downslope constrains the rates at which its subaerial salt karst terrains form. Normal seawater is only 10% saturated in NaCl. It should therefore be sufficiently aggressive to erode karst terrains into exposures of salt on the thousands of known submarine salt extrusions that have flowed or are still flowing over the floors of hundreds of submarine basins worldwide. However, we know of no attempt to constrain the processes that form submarine salt karst terrains on any of these of submarine salt extrusions. As on land, many potential submarine karst terrains are cloaked by clastic and pelagic sediments that are often hundreds of m thick. Nevertheless, detailed geophysical and bathymetric surveys have already mapped likely submarine salt karst terrains in at least the Gulf of Mexico, and the Red Sea. New images of these two areas are offered as clear evidence of submarine salt dissolution due to sinking or rising aggressive fluids. We suggest that repeated 3D surveys of distinctive features (± fixed seismic reflectors of such terrains could measure any downslope salt flow and thus offer an exceptional opportunity to constrain the rates at which submarine salt karst terrains develop. Such rates are of interest to all salt tectonicians and the many

Experiments and Modeling in Support of Generic Salt Repository Science

International Nuclear Information System (INIS)

Bourret, Suzanne Michelle; Stauffer, Philip H.; Weaver, Douglas James; Caporuscio, Florie Andre; Otto, Shawn; Boukhalfa, Hakim; Jordan, Amy B.; Chu, Shaoping; Zyvoloski, George Anthony; Johnson, Peter Jacob

2017-01-01

Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

Experiments and Modeling in Support of Generic Salt Repository Science

Energy Technology Data Exchange (ETDEWEB)

Bourret, Suzanne Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weaver, Douglas James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Caporuscio, Florie Andre [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Otto, Shawn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boukhalfa, Hakim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, Shaoping [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zyvoloski, George Anthony [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Johnson, Peter Jacob [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-19

Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

Improved Design and Fabrication of Hydrated-Salt Pills

Science.gov (United States)

Shirron, Peter J.; DiPirro, Michael J.; Canavan, Edgar R.

2011-01-01

A high-performance design, and fabrication and growth processes to implement the design, have been devised for encapsulating a hydrated salt in a container that both protects the salt and provides thermal conductance between the salt and the environment surrounding the container. The unitary salt/container structure is known in the art as a salt pill. In the original application of the present design and processes, the salt is, more specifically, a hydrated paramagnetic salt, for use as a refrigerant in a very-low-temperature adiabatic demagnetization refrigerator (ADR). The design and process can also be applied, with modifications, to other hydrated salts. Hydrated paramagnetic salts have long been used in ADRs because they have the desired magnetic properties at low temperatures. They also have some properties, disadvantageous for ADRs, that dictate the kind of enclosures in which they must be housed: Being hydrated, they lose water if exposed to less than 100-percent relative humidity. Because any dehydration compromises their magnetic properties, salts used in ADRs must be sealed in hermetic containers. Because they have relatively poor thermal conductivities in the temperature range of interest (<0.1 K), integral thermal buses are needed as means of efficiently transferring heat to and from the salts during refrigeration cycles. A thermal bus is typically made from a high-thermal-conductivity met al (such as copper or gold), and the salt is configured to make intimate thermal contact with the metal. Commonly in current practice (and in the present design), the thermal bus includes a matrix of wires or rods, and the salt is grown onto this matrix. The density and spacing of the conductors depend on the heat fluxes that must be accommodated during operation.

Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

Energy Technology Data Exchange (ETDEWEB)

Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-08-01

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

Multilayer Porous Crucibles for the High Throughput Salt Separation from Uranium Deposits

International Nuclear Information System (INIS)

Kwon, S. W.; Park, K. M.; Kim, J. G.; Kim, I. T.; Seo, B. K.; Moon, J. G.

2013-01-01

Solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as a distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while nonvolatile uranium remains behind. It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in an electro-refiner. Therefore, a wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, it was attempted to enlarge a throughput of the salt distiller with a multilayer porous crucibles for the separation of adhered salt in the uranium deposits generated from the electrorefiner. The feasibility of the porous crucibles was tested by the salt distillation experiments. In this study, the salt distiller with multilayer porous crucibles was proposed and the feasibility of liquid salt separation was examined to increase a throughput. It was found that the effective separation of salt from uranium deposits was possible by the multilayer porous crucibles

Multilayer Porous Crucibles for the High Throughput Salt Separation from Uranium Deposits

Energy Technology Data Exchange (ETDEWEB)

Kwon, S. W.; Park, K. M.; Kim, J. G.; Kim, I. T.; Seo, B. K.; Moon, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

Solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as a distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while nonvolatile uranium remains behind. It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in an electro-refiner. Therefore, a wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, it was attempted to enlarge a throughput of the salt distiller with a multilayer porous crucibles for the separation of adhered salt in the uranium deposits generated from the electrorefiner. The feasibility of the porous crucibles was tested by the salt distillation experiments. In this study, the salt distiller with multilayer porous crucibles was proposed and the feasibility of liquid salt separation was examined to increase a throughput. It was found that the effective separation of salt from uranium deposits was possible by the multilayer porous crucibles.

Effect of salt on color and warmed over flavor in charqui meat processing

Directory of Open Access Journals (Sweden)

Youssef Elza Y.

2003-01-01

Full Text Available A combination of salt (NaCl high concentration and curing salt was investigated for their role in warmed-over flavor (WOF and color changes during charqui meats processing. WOF was measured by TBARS method in uncured charqui meat (CH and in cured charqui known in Brazil as Jerked beef (JB. WOF occurred substantially in CH and sodium nitrite was able to inhibit 40-45% (p<0.05 in JB samples stored for 30 days. Color parameters also changed as evaluated by CIELAB system. The a*/b* ratio showed that CH samples presented brown color indicating the formation of metmyoglobin (Fe3+ whilst JB samples presented deep red color an indication of nitrosylmyoglobin (Fe2+ formation. Under cooking, a*/b* ratio indicated the presence of denatured metmyoglobin (Fe3+ in CH and formation of nitrosylmyochromogen (Fe2+ in JB samples. The actual iron state influenced the color of charqui meat and apparently nitrite was able to chelate Fe ions, thus inhibiting development of WOF.

Thermophysical properties of reconsolidating crushed salt.

Energy Technology Data Exchange (ETDEWEB)

Bauer, Stephen J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Urquhart, Alexander [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-03-01

Reconsolidated crushed salt is being considered as a backfilling material placed upon nuclear waste within a salt repository environment. In-depth knowledge of thermal and mechanical properties of the crushed salt as it reconsolidates is critical to thermal/mechanical modeling of the reconsolidation process. An experimental study was completed to quantitatively evaluate the thermal conductivity of reconsolidated crushed salt as a function of porosity and temperature. The crushed salt for this study came from the Waste Isolation Pilot Plant (WIPP). In this work the thermal conductivity of crushed salt with porosity ranging from 1% to 40% was determined from room temperature up to 300°C, using two different experimental methods. Thermal properties (including thermal conductivity, thermal diffusivity and specific heat) of single-crystal salt were determined for the same temperature range. The salt was observed to dewater during heating; weight loss from the dewatering was quantified. The thermal conductivity of reconsolidated crushed salt decreases with increasing porosity; conversely, thermal conductivity increases as the salt consolidates. The thermal conductivity of reconsolidated crushed salt for a given porosity decreases with increasing temperature. A simple mixture theory model is presented to predict and compare to the data developed in this study.

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

Comment and response document for the UMTRA Project vitro processing site completion report Salt Lake City, Utah. Revision 1

International Nuclear Information System (INIS)

1995-03-01

This Comment and Response Document is a series of UMTRA document review forms regarding the UMTRA Project Vitro Processing Site Completion Report for Salt Lake City, Utah in March, 1995. The completion report provides evidence that the final Salt Lake City, Utah, processing site property conditions are in accordance with the approved design and that all U.S. Environmental Protection Agency (EPA) standards have been satisfied. Included as appendices to support the stated conclusions are the record drawings; a summary of grid test results; contract specifications and construction drawings, the EPA standards (40 CFR Part 192); the audit, inspection, and surveillance summary; the permit information; and project photographs. The principal objective of the remedial action at Salt Lake City is to remove the tailings from the processing site, render the site free of contamination to EPA standards, and restore the site to the final design grade elevations. Each section is evaluated in detail to check all aspects of above report, especially the inclusion of adequate verification data. Each review form contains a section entitled State of Utah Response and Action, which is an explanation or correction of DOE criticisms of the report

Salt splitting with ceramic membranes

International Nuclear Information System (INIS)

Kurath, D.

1996-01-01

The purpose of this task is to develop ceramic membrane technologies for salt splitting of radioactively contaminated sodium salt solutions. This technology has the potential to reduce the low-level waste (LLW) disposal volume, the pH and sodium hydroxide content for subsequent processing steps, the sodium content of interstitial liquid in high-level waste (HLW) sludges, and provide sodium hydroxide free of aluminum for recycle within processing plants at the DOE complex. Potential deployment sites include Hanford, Savannah River, and Idaho National Engineering Laboratory (INEL). The technical approach consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON). As the name implies, sodium ions are transported rapidly through these ceramic crystals even at room temperatures

Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

International Nuclear Information System (INIS)

Yoshioka, Ritsuo; Mitachi, Koshi

2013-01-01

The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

Integrated membrane distillation-crystallization: process design and cost estimations for seawater treatment and fluxes of single salt solutions

NARCIS (Netherlands)

Creusen, R.J.M.; Medevoort, J. van; Roelands, C.P.M.; Renesse van Duivenbode, J.A.D. van; Hanemaaijer, J.H.; Leerdam, R.C. van

2013-01-01

The goal of this research is to design an integrated membrane distillation-crystallization (MDC) process for desalination of seawater with pure water and dry salts as the only products. The process is based on a combination of membrane distillation (MD) and osmotic distillation (OD) steps with

Residual salt separation from simulated spent nuclear fuel reduced in a LiCl-Li2O salt

International Nuclear Information System (INIS)

Hur, Jin-Mok; Hong, Sun-Seok; Seo, Chung-Seok

2006-01-01

The electrochemical reduction of spent nuclear fuel in LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel Atomic Energy Society of Japan, Tokyo, Japan, All rights reservedopyriprocess. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of the LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below the temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950degC were measured as follows: LiCl-8 wt% Li 2 O>LiCl>LiCl-8 wt% SrCl 2 >SrCl 2 . (author)

Effects of hydrologic conditions on biogeochemical processes and organic pollutant degradation in salt marsh sediments

Science.gov (United States)

W. James Catallo

2000-01-01

This work addressed the influence of tidal vs. static hydrologic conditions on biogeochemical processes and the transformation of pollutant organic chemicals (eight representative N-, O-, and S-heterocycles (NOSHs) from coal chemicals, crude oils, and pyrogenic mixtures) in salt marsh sediments. The goals were to: (1) determine the effects of static (flooded, drained)...

The effect of salt replacers and flavor enhancer on the processing characteristics and consumer acceptance of turkey sausages.

Science.gov (United States)

Pietrasik, Zeb; Gaudette, Nicole J

2015-07-01

Producing high-quality processed meats that contain reduced amounts of sodium chloride is a major challenge facing industry owing to the importance of sodium chloride toward the functional, microbial stability and sensory properties of these products. In order to create reduced sodium alternatives, a number of commercial salt replacers and flavor enhancers have entered the market; however, their ability to be applied in processed meats requires investigation. In this study, two salt replacers (Ocean's Flavor - OF45, OF60) and one flavor enhancer (Fonterra™ Savoury Powder - SP) were evaluated for their ability to effectively reduce sodium while maintaining the functional and sensory properties of turkey sausages. Functionality via instrumental measures (yield, purge loss, pH, expressible moisture, proximate composition, sodium content, color, texture), safety (microbiological assessment) and consumer acceptability were obtained on all samples. All non-control treatments resulted in products with sodium chloride contents below Canada's Health Check™ Program target for processed meats. There was no detrimental effect on water binding and texture in treatments when NaCl was substituted with OF60 sea salt replacers. Sodium reduction had no negative effect on the shelf life of the turkey sausages with up to 60 days of refrigerated storage. Consumer acceptability for all attributes did not differ significantly, except for aftertaste, which scored lowest for OF45 compared with the control (regular NaCl content). This work demonstrated that salt replacers could potentially substitute for NaCl in smoked turkey sausages; however, further flavor optimization may be required to suppress undesirable levels of bitterness elicited by some of these ingredients. © 2014 Society of Chemical Industry.

Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

International Nuclear Information System (INIS)

Hsu, P.C.

1997-01-01

Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

A novel process for recovery of iron, titanium, and vanadium from titanomagnetite concentrates: NaOH molten salt roasting and water leaching processes

International Nuclear Information System (INIS)

Chen, Desheng; Zhao, Longsheng; Liu, Yahui; Qi, Tao; Wang, Jianchong; Wang, Lina

2013-01-01

Highlights: ► The conversion of titanium is 96.6% in the rich titanium–vanadium slag. ► MgTi 2 O 5 and M 3 O 5 (M = Ti, Mg, Fe) were converted to Na 2 TiO 3 and NaMO 2 , respectively. ► Na 2 TiO 3 is converted to undefined structure of H 2 TiO 3 . ► NaMO 2 is converted to α-NaFeO 2 -type structure of HMO 2 . ► 87.3% of sodium, 42.3% of silicon, 43.2% of aluminum, 22.8% of manganese and 96.6% of vanadium were leached out. -- Abstract: A novel process for recovering iron, titanium, and vanadium from titanomagnetite concentrates has been developed. In the present paper, the treatment of rich titanium–vanadium slag by NaOH molten salt roasting and water leaching processes is investigated. In the NaOH molten salt roasting process, the metallic iron is oxidized into ferriferous oxide, MgTi 2 O 5 is converted to NaCl-type structure of Na 2 TiO 3 , and M 3 O 5 (M = Ti, Mg, Fe) is converted to α-NaFeO 2 -type structure of NaMO 2 , respectively. Roasting temperature and NaOH–slag mass ratio played a considerable role in the conversion of titanium in the rich titanium–vanadium slag during the NaOH molten salt roasting process. Roasting at 500 °C for 60 min and a 1:1 NaOH–slag mass ratio produces 96.3% titanium conversion. In the water leaching process, the Na + was exchanged with H + , Na 2 TiO 3 is converted to undefined structure of H 2 TiO 3 , and NaMO 2 is converted to α-NaFeO 2 -type structure of HMO 2 . Under the optimal conditions, 87.3% of the sodium, 42.3% of the silicon, 43.2% of the aluminum, 22.8% of the manganese, and 96.6% of the vanadium are leached out

R and D activities on the management of waste chloride salts in KAERI

International Nuclear Information System (INIS)

In-Tae, Kim; Hwan-Seo, Park; Jeong-Gook, Kim; Hee-Chul, Yang; Yong-Joon, Cho; Eung-Ho Kim

2007-01-01

Full text of publication follows. Electrochemical treatment of spent oxide fuels has been intensively studied in KAERI to reduce the volume, heat load and radiotoxicity of high-level wastes. It consists of an electrolytic reduction process to convert the oxide fuel into a metallic form and an electro-refining process to separate TRU elements from the electro-reduced metal ingot. Two types of waste salts are expected to generate from the electrochemical pyro-processes, that is, LiCl salt from the reduction process and LiCl+KCl eutectic salt form the refining process. The R and D strategy of the waste salt management in KAERI can be categorized into two parts: 1) enhancement of safety by the stabilisation/solidification of waste salt that is to be finally disposed of and 2) reduction of the waste generation by the regeneration/recycle of the spent salt after removal of radionuclides in it. A sol-gel technique and a zeolite occlusion technique are under development to stabilize the waste salt. The LiCl salt is stabilised by a low-temperature sol-gel process and then the gel product is solidified into a ceramic-like waste form with an addition of glass frit. Another method uses Zeolite-4A to occlude the LiCl salt into its cage and adsorption site to immobilize the radionuclides. The product, salt-occluded zeolite, is fabricated into another type of a ceramic waste form. For the regeneration and recycle of the spent salt, the radionuclides in the salt are removed by a zeolite process for the LiCl salt and by an oxidation/distillation process for the eutectic salt. The target nuclides to be removed in each process are Cs/Sr and rare earth (RE) elements, respectively. In the oxidation/ distillation process, the rare earth chloride nuclides are oxidised by an oxygen sparging method, and the products are precipitated in the form of oxide or oxychloride REs. After separation of the RE elements from the precipitates by distillation, the refined spent salt with a low content

Preparation and characterization of hydrated salts/silica composite as shape-stabilized phase change material via sol–gel process

International Nuclear Information System (INIS)

Wu, Yuping; Wang, Tao

2014-01-01

Highlights: • A mixture of hydrated salts were adopted as phase change materials. • Phase segregation of the hydrated salts was inhibited. • Subcooling was slightly mitigated. • Thermal cycling performance was greatly improved after PVP coating. - Abstract: A novel shape-stabilized phase change material composite was prepared by impregnating the mixture of hydrated salts (Na 2 SO 4 ·10H 2 O–Na 2 HPO 4 ·12H 2 O) into porous silica matrix obtained by sol–gel process and further coated with polyvinylpyrrolidone (PVP) to improve the thermal cycling performance. The chemical compatibility, morphology and phase change properties were investigated by Fourier transform infrared spectroscopy (FT-IR), scanning electron microscope (SEM), hot-stage polarizing optical microscope (HS-POM) and differential scanning calorimetry (DSC). Confined in the silica matrix, phase segregation of the hydrated salts was inhibited and subcooling was slightly mitigated. No leakage was observed during the solid–liquid phase transition even when the mass ratio of hydrated salts to silica was as high as 70:30. Results showed that the melting enthalpy of the composite can reach 106.2 kJ/kg with the melting temperature at 30.13 °C and there was no significant enthalpy loss after 30 thermal cycles

SEPARATION OF METAL SALTS BY ADSORPTION

Science.gov (United States)

Gruen, D.M.

1959-01-20

It has been found that certain metal salts, particularly the halides of iron, cobalt, nickel, and the actinide metals, arc readily absorbed on aluminum oxide, while certain other salts, particularly rare earth metal halides, are not so absorbed. Use is made of this discovery to separate uranium from the rare earths. The metal salts are first dissolved in a molten mixture of alkali metal nitrates, e.g., the eutectic mixture of lithium nitrate and potassium nitrate, and then the molten salt solution is contacted with alumina, either by slurrying or by passing the salt solution through an absorption tower. The process is particularly valuable for the separation of actinides from lanthanum-group rare earths.

Mechanical stratification of autochthonous salt: Implications from basin-scale numerical models of rifted margin salt tectonics

Science.gov (United States)

Ings, Steven; Albertz, Markus

2014-05-01

Deformation of salt and sediments owing to the flow of weak evaporites is a common phenomenon in sedimentary basins worldwide, and the resulting structures and thermal regimes have a significant impact on hydrocarbon exploration. Evaporite sequences ('salt') of significant thickness (e.g., >1km) are typically deposited in many cycles of seawater inundation and evaporation in restricted basins resulting in layered autochthonous evaporite packages. However, analogue and numerical models of salt tectonics typically treat salt as a homogeneous viscous material, often with properties of halite, the weakest evaporite. In this study, we present results of two-dimensional plane-strain numerical experiments designed to illustrate the effects of variable evaporite viscosity and embedded frictional-plastic ('brittle') sediment layers on the style of salt flow and associated deformation of the sedimentary overburden. Evaporite viscosity is a first-order control on salt flow rate and the style of overburden deformation. Near-complete evacuation of low-viscosity salt occurs beneath expulsion basins, whereas significant salt is trapped when viscosity is high. Embedded frictional-plastic sediment layers (with finite yield strength) partition salt flow and develop transient contractional structures (folds, thrust faults, and folded faults) in a seaward salt-squeeze flow regime. Multiple internal sediment layers reduce the overall seaward salt flow during sediment aggradation, leaving more salt behind to be re-mobilized during subsequent progradation. This produces more seaward extensive allochthonous salt sheets. If there is a density difference between the embedded layers and the surrounding salt, then the embedded layers 'fractionate' during deformation and either float to the surface or sink to the bottom (depending on density), creating a thick zone of pure halite. Such a process of 'buoyancy fractionation' may partially explain the apparent paradox of layered salt in

Numerical simulation of minor actinide recovery behaviour in batch processing of spent metallic fuel by electrorefining

Energy Technology Data Exchange (ETDEWEB)

Nawada, H P; Bhat, N P [Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Balasubramanian, G R [Atomic Energy Commission, Mumbai (India)

1994-06-01

Numerical simulation of electro-transport of fuel actinides (FAs), minor actinides (MAs) and rare earths (REs) in the electro-refiner (ER) for pyrochemical reprocessing of a typical spent IFR metallic fuel has been attempted based on improved thermo-chemical model developed for application to multi-component system in the ER. Optimization of MA recovery and decontamination factors (DFs) for MAs and REs in batch processing is presented. (author). 7 refs., 4 figs., 1 tab.

Advances in molten salt electrochemistry towards future energy systems

International Nuclear Information System (INIS)

Ito, Yasuhiko

2005-01-01

This review article describes some selected novel molten salt electrochemical processes which have been created/developed by the author and his coworkers, with emphasis on the applications towards future energy systems. After showing a perspective of the applications of molten salt electrochemistry from the viewpoints of energy and environment, several selected topics are described in detail, which include nitride fuel cycle in a nuclear field, hydrogen energy system coupled with ammonia economy, thermally regenerative fuel cell systems, novel Si production process for solar cell and novel molten salt electrochemical processes for various energy and environment related functional materials including nitrides, rare earth-transition metal alloys, fine particles obtained by plasma-induced electrolysis, and carbon film. And finally, the author stresses again, the importance and potential of molten salt electrochemistry, and encourages young students, scientists and researchers to march in a procession hand in hand towards a bright future of molten salts. (author)

Integration of membrane distillation into traditional salt farming method: Process development and modelling

Science.gov (United States)

Hizam, S.; Bilad, M. R.; Putra, Z. A.

2017-10-01

Farmers still practice the traditional salt farming in many regions, particularly in Indonesia. This archaic method not only produces low yield and poor salt quality, it is also laborious. Furthermore, the farming locations typically have poor access to fresh water and are far away from electricity grid, which restrict upgrade to a more advanced technology for salt production. This paper proposes a new concept of salt harvesting method that improves the salt yield and at the same time facilitates recovery of fresh water from seawater. The new concept integrates solar powered membrane distillation (MD) and photovoltaic cells to drive the pumping. We performed basic solar still experiments to quantify the heat flux received by a pond. The data were used as insight for designing the proposed concept, particularly on operational strategy and the most effective way to integrate MD. After the conceptual design had been developed, we formulated mass and energy balance to estimate the performance of the proposed concept. Based on our data and design, it is expected that the system would improve the yield and quality of the salt production, maximizing fresh water harvesting, and eventually provides economical gain for salt farmers hence improving their quality of life. The key performance can only be measured via experiment using gain output ratio as performance indicator, which will be done in a future study.

Salt geologic evaluation of the impact of cryogenic fissures and halokinetic deformation processes on the integrity of the geological barrier of the salt dome Gorleben

International Nuclear Information System (INIS)

Hammer, Joerg; Fleig, Stephanie; Mingerzahn, Gerhard

2012-07-01

In several salt domes of the area close to Hannover fissures were observed that might be caused by thermally induced fissure formation due to cold periods (cryogenic fissures). Comprehensive substantial-structural analyses are performed as an example for the salt dome Bokeloh with respect to genesis and transferability to the salt dome Gorleben. Based on recent structure-geological, mineralogical-geochemical and micro-paleontological studies and thermo-mechanical modeling a solely thermally induced fissure formation due to cold periods is unlikely for the salt dome Bokeloh. There is a direct relation between the genesis of the salt dome Bokeloh, its regional tectonic site and the fissure formation. Due to the completely different genesis and another regional-tectonic situation the existence of cryogenic fissures is excluded for the salt dome Gorleben. The salt-geologic and experimental studies on the deformation of anhydrite layers in salt domes are summarized and evaluated with respect to the long-term consequences for a potential final repository for high-level heat-generating radioactive waste in the salt dome Gorleben. The studies confirm the older BGR studies that anhydrite layers do not represent hydraulic potential ling-distance liquid paths.

LIFE Materails: Molten-Salt Fuels Volume 8

Energy Technology Data Exchange (ETDEWEB)

Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P

2008-12-11

The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

LIFE Materails: Molten-Salt Fuels Volume 8

International Nuclear Information System (INIS)

Moir, R.; Brown, N.; Caro, A.; Farmer, J.; Halsey, W.; Kaufman, L.; Kramer, K.; Latkowski, J.; Powers, J.; Shaw, H.; Turchi, P.

2008-01-01

The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

Salt at concentrations relevant to meat processing enhances Shiga toxin 2 production in Escherichia coli O157:H7.

Science.gov (United States)

Harris, Shaun M; Yue, Wan-Fu; Olsen, Sarena A; Hu, Jia; Means, Warrie J; McCormick, Richard J; Du, Min; Zhu, Mei-Jun

2012-10-15

Escherichia coli (E. coli) O157:H7 remains a major food safety concern associated with meat, especially beef products. Shiga toxins (Stx) are key virulence factors produced by E. coli O157:H7 that are responsible for hemorrhagic colitis and Hemolytic Uremic Syndrome. Stx are heat stable and can be absorbed after oral ingestion. Despite the extensive study of E. coli O157:H7 survival during meat processing, little attention is paid to the production of Stx during meat processing. The objective of this study was to elucidate the effect of salt, an essential additive to processed meat, at concentrations relevant to meat processing (0%, 1%, 2%, 3%, W/V) on Stx2 production and Stx2 prophage induction by E. coli O157:H7 strains. For both E. coli O157:H7 86-24 and EDL933 strains, including 2% salt in LB broth decreased (Pmeat processing enhances Stx production, a process linked to bacterial stress response and lambdoid prophage induction. Published by Elsevier B.V.

Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

International Nuclear Information System (INIS)

Nuttin, Alexis

2002-01-01

In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232 Th/ 233 U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

A systematic technique for the sequential restoration of salt structures

Science.gov (United States)

Rowan, Mark G.

1993-12-01

A method is described for the sequential restoration of cross sections in areas of salt tectonics where deformation is confined to the salt and higher layers. The subsurface geometry evolves with time through the interaction of various processes: sedimentation, compaction, isostatic adjustment, thermal subsidence (if present), faulting, and salt withdrawal/ diapirism. The technique systematically calculates and removes the effects of each of these processes during specified time intervals defined by the interpreted horizons. It makes no assumptions about salt kinematics and generally results in the area of the salt layer changing through time. The method is described for restoration of extensional terranes, but it is also suitable for areas of contractional salt tectonics with only minor modifications. After converting an interpreted seismic profile to depth, the top layer is stripped off and the underlying section is decompacted according to standard porosity-depth functions. A deep baseline, unaffected by compaction or deformation, is used to restore any isostatic compensation or thermal subsidence. Isostasy is calculated according to the Airy model, and differential sedimentary loading across a section is shown to be approximately balanced by changes in salt thickness so that the load is evenly distributed. After these processes have been reversed, the resulting geometry and the seismic data are used to create the sea-floor template for structural restoration. Fault offsets are removed and the layers down to the top salt are restored to this template, while the base salt remains fixed. The resulting space between the restored top salt and the fixed base salt defines the restored salt geometry. In addition, the difference between the sea-floor template and a fixed sea level provides a measure of the change in water depth (ignoring eustatic changes in sea level). The technique is applied to an interpreted seismic profile from the eastern Green Canyon/Ewing Bank

Accelerator-driven molten-salt blankets: Physics issues

International Nuclear Information System (INIS)

Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

1994-01-01

A number of nuclear physics issues concerning the Los Alamos molten-salt, accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external, moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

Accelerator-driven molten-salt blankets: Physics issues

International Nuclear Information System (INIS)

Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

1994-01-01

A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

Remote engineering progress report, January-December 1982

International Nuclear Information System (INIS)

Brown, C.M.

1984-01-01

Discussed briefly are the objectives, prior work (if any), present achievements, and future work in the following areas: an automated storage, transfer, and retrieval system for laboratory specimens; automation of molten salt button breakout operations; emission spectrometer automation, direct oxide reduction process support; hydride operations support; part serialization processes; contact handled waste container welding; automated sample cutting; filter probe positioner; Unimate robot replacement to handle uranium ingots in a hot pressing operation; development and testing of a device to automate compacting on site return nuclear weapons parts for pyrochemical operations; remote transfer system for gloveboxes (made from a modified radio-controlled toy tank); plutonium oxide/skull burn and reburn process; advanced size reduction facility; automation of the plutonium oxide pelletizing process for chemical analytical laboratory operations; automated assembly demonstration for a Mechanical Safe and Arm Detonator (MSAD); and a stacker-retriever remote vacuum cleaning system

Salt disposal of heat-generating nuclear waste

International Nuclear Information System (INIS)

Leigh, Christi D.; Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

Salt disposal of heat-generating nuclear waste.

Energy Technology Data Exchange (ETDEWEB)

Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

Liquid Salt Heat Exchanger Technology for VHTR Based Applications

Energy Technology Data Exchange (ETDEWEB)

Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

2012-10-11

The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

Development of High-Temperature Transport System for Molten Salt in Pyroprocessing

International Nuclear Information System (INIS)

Lee, Sung Ho; Kim, In Tae; Park, Sung Bin

2014-01-01

The electrorefining process, which is a key process in pyroprocessing, is composed of two parts, electrorefining to deposit a uranium with a solid cathode and electrowinning to co-deposit TRU and RE with a liquid cadmium cathode (LCC). As the electrorefining operation proceedes, TRU and RE are accumulated in electrolyte LiCl-KCl salt, and after the electrorefining process, the molten salt used in an electrorefining reactor should by transported to the next process, the electrowinning process, to recover U/TRU/RE; Thus, a molten salt transfer system by suction is now being developed. An apparatus for suction transport experiments was designed and constructed for the development of high- temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt. The feasibility of pyro-reprocessing has been demonstrated through many laboratory-scale experiments. In pyroprocessing, a eutectic LiCl-KCl salt was used as a liquid elextrolyte for a recovery of actinides. However, reliable transport technologies for these high temperature liquids have not yet been developed. A preliminary study on high-temperature transport technology for molten salt by suction is now being carried out. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated to select the most suitable method for molten salt transport. An apparatus for suction transport experiments was designed and installed for the development of high-temperature molten salt transport technology. Basic preliminary suction transport experiments were carried out using the prepared LiC-KCl eutectic salt at 500 .deg. C to observe the transport behavior of LiCl-KCl molten salt. In addition, a PRIDE salt transport system was designed and installed for an engineering-scale salt transport demonstration. Several types of suction transport experiments using molten salt (LiCl-KCl eutectics) for the development of a high

Development of High-Temperature Transport System for Molten Salt in Pyroprocessing

Energy Technology Data Exchange (ETDEWEB)

Lee, Sung Ho; Kim, In Tae; Park, Sung Bin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The electrorefining process, which is a key process in pyroprocessing, is composed of two parts, electrorefining to deposit a uranium with a solid cathode and electrowinning to co-deposit TRU and RE with a liquid cadmium cathode (LCC). As the electrorefining operation proceedes, TRU and RE are accumulated in electrolyte LiCl-KCl salt, and after the electrorefining process, the molten salt used in an electrorefining reactor should by transported to the next process, the electrowinning process, to recover U/TRU/RE; Thus, a molten salt transfer system by suction is now being developed. An apparatus for suction transport experiments was designed and constructed for the development of high- temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt. The feasibility of pyro-reprocessing has been demonstrated through many laboratory-scale experiments. In pyroprocessing, a eutectic LiCl-KCl salt was used as a liquid elextrolyte for a recovery of actinides. However, reliable transport technologies for these high temperature liquids have not yet been developed. A preliminary study on high-temperature transport technology for molten salt by suction is now being carried out. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated to select the most suitable method for molten salt transport. An apparatus for suction transport experiments was designed and installed for the development of high-temperature molten salt transport technology. Basic preliminary suction transport experiments were carried out using the prepared LiC-KCl eutectic salt at 500 .deg. C to observe the transport behavior of LiCl-KCl molten salt. In addition, a PRIDE salt transport system was designed and installed for an engineering-scale salt transport demonstration. Several types of suction transport experiments using molten salt (LiCl-KCl eutectics) for the development of a high

An application of LOTEM around salt dome near Houston, Texas

Science.gov (United States)

Paembonan, Andri Yadi; Arjwech, Rungroj; Davydycheva, Sofia; Smirnov, Maxim; Strack, Kurt M.

2017-07-01

A salt dome is an important large geologic structure for hydrocarbon exploration. It may seal a porous reservoir of rocks that form petroleum reservoirs. Several techniques such as seismic, gravity, and electromagnetic including magnetotelluric have successfully yielded salt dome interpretation. Seismic has difficulties seeing through the salt because the seismic energy gets trapped by the salt due to its high velocity. Gravity and electromagnetics are more ideal methods. Long Offset Transient Electromagnetic (LOTEM) and Focused Source Electromagnetic (FSEM) were tested over a salt dome near Houston, Texas. LOTEM data were recorded at several stations with varying offset, and the FSEM tests were also made at some receiver locations near a suspected salt overhang. The data were processed using KMS's processing software: First, for assurance, including calibration and header checking; then transmitter and receiver data are merged and microseismic data is separated; Finally, data analysis and processing follows. LOTEM processing leads to inversion or in the FSEM case 3D modeling. Various 3D models verify the sensitivity under the salt dome. In addition, the processing was conducted pre-stack, stack, and post-stack. After pre-stacking, the noise was reduced, but showed the ringing effect due to a low-pass filter. Stacking and post-stacking with applying recursive average could reduce the Gibbs effect and produce smooth data.

Brine Transport Experiments in Granular Salt

Energy Technology Data Exchange (ETDEWEB)

Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boukhalfa, Hakim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Caporuscio, Florie Andre [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-06-06

To gain confidence in the predictive capability of numerical models, experimental validation must be performed to ensure that parameters and processes are correctly simulated. The laboratory investigations presented herein aim to address knowledge gaps for heat-generating nuclear waste (HGNW) disposal in bedded salt that remain after examination of prior field and laboratory test data. Primarily, we are interested in better constraining the thermal, hydrological, and physicochemical behavior of brine, water vapor, and salt when moist salt is heated. The target of this work is to use run-of-mine (RoM) salt; however during FY2015 progress was made using high-purity, granular sodium chloride.

Thermodynamic analysis of salt corrosion of titanium alloys

International Nuclear Information System (INIS)

Travkin, V.V.; Pshirkov, V.F.; Kolachev, B.A.

1979-01-01

About 200 possible chemical reactions of metals, salts and oxides (in a solid state) with water (in a vapour state), and with gases (O 2 , Cl 2 , HCl) were studied by the thermodynamic analysis to elucidate a chemical nature of processes taking place at salt corrosion of titanium alloys (VT22, VT6 and VT16). Temperature dependences of isobaric-isothermic potential were considered to reveal a possibility of spontaneous course and direction of reactions as well as to obtain a comparative estimate of the probability of their pro-cedure. Thermodynamically possible schemes of the chemism of titanium alloy salt corrosion are proposed. Complex che-mical reactions take place in the presence of salt, moisture and oxygen of air on the surface of the alloys. The reactions proceed with the formation of titanium and alloying component chlorides, free chlorine and hydrogen. The free chlorine or HCl are released during pyrohydrolysis and oxidation of chlo-rides. The former ones interact with the alloy with the formation of salts, and hydrogen may be absorbed by the metal and cause embrittlement. Chlorides on the metal surface accelerate the chlorination process. NaCl acts as a cata-lyst. The determination of salt corrosion products has confirmed the process mechanism proposed

Molten salt extraction process for the recovery of valued transition metals from land-based and deep-sea minerals

Science.gov (United States)

Maroni, V.A.; von Winbush, S.

1987-05-01

A process for extracting transition metals and particularly cobalt and manganese together with iron, copper and nickel from low grade ores (including ocean-floor nodules) by converting the metal oxides or other compositions to chlorides in a molten salt, and subsequently using a combination of selective distillation at temperatures below about 500/degree/C, electrolysis at a voltage not more negative that about /minus/1.5 volt versus Ag/AgCl, and precipitation to separate the desired manganese and cobalt salts from other metals and provide cobalt and manganese in metallic forms or compositions from which these metals may be more easily recovered.

Engineering development studies for molten-salt breeder reactor processing No. 18

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1975-03-01

A water--mercury system was used to study the effect of geometric variations on mass transfer rates in rectangular contractors similar to those proposed for the molten-salt breeder reactor (MSBR) fuel reprocessing scheme. Since mass transfer rates were not accurately predicted by the Lewis correlation, other correlations were investigated. A correlation which was found to fit the experimental results is given. Mass transfer rates are being measured in a fluoride salt--bismuth contactor. Experimental results indicate that the mass transfer rates in the salt--bismuth system fall between the Lewis correlation and the modified correlation given above. Autoresistance heating tests were continued in the fluorinator mock-up using LiF--BeF 2 --ThF 4 (72-16-12 mole percent) salt. The equipment was returned to operating condition, and five experiments were run. Although correct steady-state operation was not achieved, the results were encouraging. A two-dimensional electrical analog was constructed to study current flow through the electrode sidearm and other critical areas of the test vessel. These studies indicate that no regions of abnormally high current density existed in the first nine runs with the present autoresistance heating equipment. Localized heating had previously been the suspected cause for the failure to achieve proper operation of this equipment. (U.S.)

Nanoscopic characterization of the water vapor-salt interfacial layer reveals a unique biphasic adsorption process

Science.gov (United States)

Yang, Liu; He, Jianfeng; Shen, Yi; Li, Xiaowei; Sun, Jielin; Czajkowsky, Daniel M.; Shao, Zhifeng

2016-08-01

Our quantitative understanding of water adsorption onto salt surfaces under ambient conditions is presently quite poor owing to the difficulties in directly characterizing this interfacial layer under these conditions. Here we determine the thickness of the interfacial layer on NaCl at different relative humidities (RH) based on a novel application of atomic force spectroscopy and capillary condensation theory. In particular, we take advantage of the microsecond-timescale of the capillary condensation process to directly resolve the magnitude of its contribution in the tip-sample interaction, from which the interfacial water thickness is determined. Further, to correlate this thickness with salt dissolution, we also measure surface conductance under similar conditions. We find that below 30% RH, there is essentially only the deposition of water molecules onto this surface, typical of conventional adsorption onto solid surfaces. However, above 30% RH, adsorption is simultaneous with the dissolution of ions, unlike conventional adsorption, leading to a rapid increase of surface conductance. Thus, water adsorption on NaCl is an unconventional biphasic process in which the interfacial layer not only exhibits quantitative differences in thickness but also qualitative differences in composition.

Applying commercial robotic technology to radioactive material processing

International Nuclear Information System (INIS)

Grasz, E.L.; Sievers, R.H. Jr.

1991-01-01

The development of robotic systems to automate nuclear material processing in glove boxes is motivated by the need to reduce operator radiation exposure, minimize the generation of process waste, and to improve security of nuclear materials. Commercial robotic systems can furnish the needed manipulation capabilities by are not readily compatible with the glove box environment and physical restrictions. Alpha radiation, concentrated dust, a dry atmosphere and restricted work space require unique adaptations of commercial robotics. Tradeoffs between meeting desired functional capabilities and extensive customization are necessary. The reported design and development efforts include evaluating the feasibility of using a commercial gantry robot for glove box pyrochemical and smelting operations. This paper presents the initial results and observations for this development effort

Mechanism of Process-Induced Salt-to-Free Base Transformation of Pharmaceutical Products

DEFF Research Database (Denmark)

Bruun Hansen, Thomas; Qu, Haiyan

2014-01-01

pH-solubility profiles of a model drug in salt form was established and the mechanism of salt-to-free base form transformation was investigated by increasing pH of the system. Wet massing experiments along with suspension experiments were used to investigate the effects of excipients on the stabi...

Salt-induced root protein profile changes in seedlings of maize inbred lines with differing salt tolerances

Directory of Open Access Journals (Sweden)

Yujing Cheng

2014-12-01

Full Text Available Salt stress is one of the severest growth limited-factors to agriculture production. To gain in-depth knowledge of salt-stress response mechanisms, the proteomics analysis from two maize (Zea mays L. inbred lines was carried out using two-dimensional gel electrophoresis (2-DGE and matrix-assisted laser desorption/ionization time-of-flight mass spectrometry (MALDI-TOF/TOF-MS. There were 57 salt-regulated proteins identified, 21 and 36 proteins were differentially regulated in inbred lines 'Nongda 1145' (salt-resistant and 'D340' (salt-sensitive, respectively. The identified proteins were distributed in 11 biological processes and seven molecular functions. Under salt stress, proteins related to antioxidation and lignin synthesis were increased in both inbred lines. The relative abundance of proteins involved in translation initiation, elongation, and protein proteolysis increased in 'Nongda 1145' and decreased in 'D340'. In addition, the abundance of proteins involved in carbohydrate metabolism, protein refolding, ATP synthase and transcription differed between the two inbred lines. Our results suggest that the enhanced ability of salt-tolerant inbred line 'Nongda 1145' to combat salt stress occurs via regulation of transcription factors promoting increased antioxidation and lignin biosynthesis, enhanced energy production, and acceleration of protein translation and protein proteolysis.

Seismic anisotropy in deforming salt bodies

Science.gov (United States)

Prasse, P.; Wookey, J. M.; Kendall, J. M.; Dutko, M.

2017-12-01

Salt is often involved in forming hydrocarbon traps. Studying salt dynamics and the deformation processes is important for the exploration industry. We have performed numerical texture simulations of single halite crystals deformed by simple shear and axial extension using the visco-plastic self consistent approach (VPSC). A methodology from subduction studies to estimate strain in a geodynamic simulation is applied to a complex high-resolution salt diapir model. The salt diapir deformation is modelled with the ELFEN software by our industrial partner Rockfield, which is based on a finite-element code. High strain areas at the bottom of the head-like strctures of the salt diapir show high amount of seismic anisotropy due to LPO development of halite crystals. The results demonstrate that a significant degree of seismic anisotropy can be generated, validating the view that this should be accounted for in the treatment of seismic data in, for example, salt diapir settings.