Analysis of cadmium in undissolved anode materials of Mark-IV electro-refiner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoo, Tae-Sic; Fredrickson, G.L.; Vaden, D.

2013-07-01

The Mark-IV electro-refiner (Mk-IV ER) is a unit process in the FCF (Fuel Conditioning Facility), which is primarily assigned to treating the used driver fuels. Mk-IV ER contains an electrolyte/molten cadmium system for refining uranium electrochemically. Typically, the anode of the Mk-IV ER consists of the chopped sodium-bonded metallic driver fuels, which have been primarily U-10Zr binary fuels. Chemical analysis of the residual anode materials after electrorefining indicates that a small amount of cadmium is removed from the Mk-IV ER along with the undissolved anode materials. Investigation of chemical analysis data indicates that the amount of cadmium in the undissolvedmore » anode materials is strongly correlated with the anode rotation speeds and the residence time of the anode in the Mk-IV ER. Discussions are given to explain the prescribed correlation. (authors)« less

Universal fuel basket for use with an improved oxide reduction vessel and electrorefiner vessel

DOEpatents

Herrmann, Steven D.; Mariani, Robert D.

2002-01-01

A basket, for use in the reduction of UO.sub.2 to uranium metal and in the electrorefining of uranium metal, having a continuous annulus between inner and outer perforated cylindrical walls, with a screen adjacent to each wall. A substantially solid bottom and top plate enclose the continuous annulus defining a fuel bed. A plurality of scrapers are mounted adjacent to the outer wall extending longitudinally thereof, and there is a mechanism enabling the basket to be transported remotely.

Advanced electrorefiner design

DOEpatents

Miller, W.E.; Gay, E.C.; Tomczuk, Z.

1996-07-02

A combination anode and cathode is described for an electrorefiner which includes a hollow cathode and an anode positioned inside the hollow cathode such that a portion of the anode is near the cathode. A retaining member is positioned at the bottom of the cathode. Mechanism is included for providing relative movement between the anode and the cathode during deposition of metal on the inside surface of the cathode during operation of the electrorefiner to refine spent nuclear fuel. A method is also disclosed which includes electrical power means selectively connectable to the anode and the hollow cathode for providing electrical power to the cell components, electrically transferring uranium values and plutonium values from the anode to the electrolyte, and electrolytically depositing substantially pure uranium on the hollow cathode. Uranium and plutonium are deposited at a liquid cathode together after the PuCl{sub 3} to UCl{sub 3} ratio is greater than 2:1. Slots in the hollow cathode provides close anode access for the liquid pool in the liquid cathode. 6 figs.

Advanced electrorefiner design

DOEpatents

Miller, William E.; Gay, Eddie C.; Tomczuk, Zygmunt

1996-01-01

A combination anode and cathode for an electrorefiner which includes a hollow cathode and an anode positioned inside the hollow cathode such that a portion of the anode is near the cathode. A retaining member is positioned at the bottom of the cathode. Mechanism is included for providing relative movement between the anode and the cathode during deposition of metal on the inside surface of the cathode during operation of the electrorefiner to refine spent nuclear fuel. A method is also disclosed which includes electrical power means selectively connectable to the anode and the hollow cathode for providing electrical power to the cell components, electrically transferring uranium values and plutonium values from the anode to the electrolyte, and electrolytically depositing substantially pure uranium on the hollow cathode. Uranium and plutonium are deposited at a liquid cathode together after the PuCl.sub.3 to UCl.sub.3 ratio is greater than 2:1. Slots in the hollow cathode provides close anode access for the liquid pool in the liquid cathode.

Effect of cathode material on the electrorefining of U in LiCl-KCl molten salts

NASA Astrophysics Data System (ADS)

Lee, Chang Hwa; Kim, Tack-Jin; Park, Sungbin; Lee, Sung-Jai; Paek, Seung-Woo; Ahn, Do-Hee; Cho, Sung-Ki

2017-05-01

The influence of cathode materials on the U electrorefining process is examined using electrochemical measurements and SEM-EDX observations. Stainless steel (STS), Mo, and W electrodes exhibit similar U reduction/oxidation behavior in 500 °C LiCl-KCl-UCl3 molten salts, as revealed by the cyclic voltammograms. However, slight shifts are observed in the cathodic and anodic peak potentials at the STS electrode, which are related to the fast reduction/oxidation kinetics associated with this electrode. The U deposits on the Mo and W electrodes consist of uniform dendritic chains of U in rhomboidal-shaped crystals, whereas several U dendrites protruding from the surface are observed for the STS electrode. EDX mapping of the electrode surfaces reveals that simple scraping of the U dendrites from W electrodes pretreated in dilute HCl solutions to dissolve the residual salt, results in clear removal of the U deposits, whereas a thick U deposit layer strongly adheres to the STS electrode surface even after treatment. This result is expected to contribute to the development of an effective and continuous U recovery process using electrorefining.

PLUTONIUM ELECTROREFINING CELLS

DOEpatents

Mullins, L.J. Jr.; Leary, J.A.; Bjorklund, C.W.; Maraman, W.J.

1963-07-16

Electrorefining cells for obtaining 99.98% plutonium are described. The cells consist of an impure liquid plutonium anode, a molten PuCl/sub 3/-- alkali or alkaline earth metal chloanode, a molten PuCl/sub 3/-alkali or alkaline earth metal chloride electrolyte, and a nonreactive cathode, all being contained in nonreactive ceramic containers which separate anode from cathode by a short distance and define a gap for the collection of the purified liquid plutonium deposited on the cathode. Important features of these cells are the addition of stirrer blades on the anode lead and a large cathode surface to insure a low current density. (AEC)

Process to remove rare earth from IFR electrolyte

DOEpatents

Ackerman, John P.; Johnson, Terry R.

1994-01-01

The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

Process to remove rare earth from IFR electrolyte

DOEpatents

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

1992-01-01

The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

Process to remove rare earth from IFR electrolyte

DOEpatents

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

1994-08-09

The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner. 1 fig.

Retrieving Historical Electrorefining Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, Meagan Daniella

Pyrochemical Operations began at Los Alamos National Laboratory (LANL) during 1962 (1). Electrorefining (ER) has been implemented as a routine process since the 1980’s. The process data that went through the ER operation was recorded but had never been logged in an online database. Without a database new staff members are hindered in their work by the lack of information. To combat the issue a database in Access was created to collect the historical data. The years from 2000 onward were entered and queries were created to analyze trends. These trends will aid engineering and operations staff to reach optimalmore » performance for the startup of the new lines.« less

Electrorefiner system for recovering purified metal from impure nuclear feed material

DOEpatents

Berger, John F.; Williamson, Mark A.; Wiedmeyer, Stanley G.; Willit, James L.; Barnes, Laurel A.; Blaskovitz, Robert J.

2015-10-06

An electrorefiner system according to a non-limiting embodiment of the present invention may include a vessel configured to maintain a molten salt electrolyte and configured to receive a plurality of alternately arranged cathode and anode assemblies. The anode assemblies are configured to hold an impure nuclear feed material. Upon application of the power system, the impure nuclear feed material is anodically dissolved and a purified metal is deposited on the cathode rods of the cathode assemblies. A scraper is configured to dislodge the purified metal deposited on the cathode rods. A conveyor system is disposed at a bottom of the vessel and configured to remove the dislodged purified metal from the vessel.

Experimental Studies of the Effects of Anode Composition and Process Parameters on Anode Slime Adhesion and Cathode Copper Purity by Performing Copper Electrorefining in a Pilot-Scale Cell

NASA Astrophysics Data System (ADS)

Zeng, Weizhi; Wang, Shijie; Free, Michael L.

2016-10-01

Copper electrorefining tests were conducted in a pilot-scale cell under commercial tankhouse environment to study the effects of anode compositions, current density, cathode blank width, and flow rate on anode slime behavior and cathode copper purity. Three different types of anodes (high, mid, and low impurity levels) were used in the tests and were analyzed under SEM/EDS. The harvested copper cathodes were weighed and analyzed for impurities concentrations using DC Arc. The adhered slimes and released slimes were collected, weighed, and analyzed for compositions using ICP. It was shown that the lead-to-arsenic ratio in the anodes affects the sintering and coalescence of slime particles. High current density condition can improve anode slime adhesion and cathode purity by intensifying slime particles' coalescence and dissolving part of the particles. Wide cathode blanks can raise the anodic current densities significantly and result in massive release of large slime particle aggregates, which are not likely to contaminate the cathode copper. Low flow rate can cause anode passivation and increase local temperatures in front of the anode, which leads to very intense sintering and coalescence of slime particles. The results and analyses of the tests present potential solutions for industrial copper electrorefining process.

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; D. Vaden; S.X. Li

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontallymore » displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.« less

Electrochemical concentration measurements for multianalyte mixtures in simulated electrorefiner salt

NASA Astrophysics Data System (ADS)

Rappleye, Devin Spencer

The development of electroanalytical techniques in multianalyte molten salt mixtures, such as those found in used nuclear fuel electrorefiners, would enable in situ, real-time concentration measurements. Such measurements are beneficial for process monitoring, optimization and control, as well as for international safeguards and nuclear material accountancy. Electroanalytical work in molten salts has been limited to single-analyte mixtures with a few exceptions. This work builds upon the knowledge of molten salt electrochemistry by performing electrochemical measurements on molten eutectic LiCl-KCl salt mixture containing two analytes, developing techniques for quantitatively analyzing the measured signals even with an additional signal from another analyte, correlating signals to concentration and identifying improvements in experimental and analytical methodologies. (Abstract shortened by ProQuest.).

Galvanic reduction of uranium(III) chloride from LiCl-KCl eutectic salt using gadolinium metal

NASA Astrophysics Data System (ADS)

Bagri, Prashant; Zhang, Chao; Simpson, Michael F.

2017-09-01

The drawdown of actinides is an important unit operation to enable the recycling of electrorefiner salt and minimization of waste. A new method for the drawdown of actinide chlorides from LiCl-KCl molten salt has been demonstrated here. Using the galvanic interaction between the Gd/Gd(III) and U/U(III) redox reactions, it is shown that UCl3 concentration in eutectic LiCl-KCl can be reduced from 8.06 wt.% (1.39 mol %) to 0.72 wt.% (0.12 mol %) in about an hour via plating U metal onto a steel basket. This is a simple process for returning actinides to the electrorefiner and minimizing their loss to the salt waste stream.

Nuclear fuel electrorefiner

DOEpatents

Ahluwalia, Rajesh K.; Hua, Thanh Q.

2004-02-10

The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, J.P.; Miller, W.E.

1987-11-05

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

Bus bar electrical feedthrough for electrorefiner system

DOEpatents

Williamson, Mark; Wiedmeyer, Stanley G; Willit, James L; Barnes, Laurel A; Blaskovitz, Robert J

2013-12-03

A bus bar electrical feedthrough for an electrorefiner system may include a retaining plate, electrical isolator, and/or contact block. The retaining plate may include a central opening. The electrical isolator may include a top portion, a base portion, and a slot extending through the top and base portions. The top portion of the electrical isolator may be configured to extend through the central opening of the retaining plate. The contact block may include an upper section, a lower section, and a ridge separating the upper and lower sections. The upper section of the contact block may be configured to extend through the slot of the electrical isolator and the central opening of the retaining plate. Accordingly, relatively high electrical currents may be transferred into a glovebox or hot-cell facility at a relatively low cost and higher amperage capacity without sacrificing atmosphere integrity.

Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF-KF-ZrF4 molten fluorides

NASA Astrophysics Data System (ADS)

Park, Kyoung Tae; Lee, Tae Hyuk; Jo, Nam Chan; Nersisyan, Hayk H.; Chun, Byong Sun; Lee, Hyuk Hee; Lee, Jong Hyeon

2013-05-01

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr-1Nb-1Sn-0.1Fe) were used as anode materials and electrorefining was performed in a LiF-KF eutectic 6 wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.

Isolation of Copper from a 5-Cent Coin: An Example of Electrorefining

ERIC Educational Resources Information Center

Sogo, Steven G.

2004-01-01

Copper is isolated from a 5-cent coin with the help of electrolysis. This experiment is useful for conceptual understanding of the significance of reduction potentials in situation of competition for electrons.

Solution-derived sodalite made with Si- and Ge-ethoxide precursors for immobilizing electrorefiner salt

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Lepry, William C.; Crum, Jarrod V.

2016-01-01

Chlorosodalite has the general form of Na8(AlSiO4)6Cl2 and this paper describes experiments conducted to synthesize sodalite with a solution-based approach to immobilize a simulated spent electrorefiner salt solution containing a mixture of alkali, alkaline earth, and lanthanide chlorides. The reactants used were the salt solution, NaAlO2, and either Si(OC2H5)4 or Ge(OC2H5)4. Additionally, seven different glass sintering aids (at loadings of 5 mass%) were evaluated as sintering aids for consolidating the as-made powders using a cold-press-and-sinter technique. This process of using alkoxide additives for the Group IV component can be used to produce large quantities of sodalite at near-room temperature as compared to a method where colloidal silica was used as the silica source. However, the small particle sizes inhibited densification during heat treatments.

Experimental Study of Codeposition Electrochemistry Using Mixtures of ScCl 3 and YCl 3 in LiCl-KCl Eutectic Salt at 500°C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaltry, Michael R.; Yoo, Tae-Sic; Fredrickson, Guy L.

2017-09-12

Cyclic voltammetry and chronopotentiometry tests were applied to molten LiCl-KCl eutectic at 500 °C including amounts of ScCl 3 and YCl 3. The purpose of the testing was to observe the effect of applied electrical current on the codeposition of scandium and yttrium, which were chosen as surrogate elements for uranium and plutonium, respectively. Features of the work were to vary the concentration of ScCl 3 (at relatively low concentrations) as well as varying the applied current, all with a fixed concentration of YCl 3. Results of the experiments could provide insight of uranium electrorefining and may provide evidence, whichmore » suggests the electrorefiner could be operated at lower UCl 3 concentration whereby codeposition (U and Pu) could be more effectively controlled.« less

Electrolytic systems and methods for making metal halides and refining metals

DOEpatents

Holland, Justin M.; Cecala, David M.

2015-05-26

Disclosed are electrochemical cells and methods for producing a halide of a non-alkali metal and for electrorefining the halide. The systems typically involve an electrochemical cell having a cathode structure configured for dissolving a hydrogen halide that forms the halide into a molten salt of the halogen and an alkali metal. Typically a direct current voltage is applied across the cathode and an anode that is fabricated with the non-alkali metal such that the halide of the non-alkali metal is formed adjacent the anode. Electrorefining cells and methods involve applying a direct current voltage across the anode where the halide of the non-alkali metal is formed and the cathode where the non-alkali metal is electro-deposited. In a representative embodiment the halogen is chlorine, the alkali metal is lithium and the non-alkali metal is uranium.

Solution-derived sodalite made with Si- and Ge-ethoxide precursors for immobilizing electrorefiner salt

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Lepry, William C.; Crum, Jarrod V.

Chlorosodalite has the general form of Na8(AlSiO4)6Cl2 and this paper describes experiments conducted to synthesize sodalite to immobilize a mixed chloride salt using solution-based techniques. Sodalites were made using different Group IV contributions from either Si(OC2H5)4 or Ge(OC2H5)4, NaAlO2, and a simulated spent electrorefiner salt solution containing a mixture of alkali, alkaline earth, and lanthanide chlorides. Additionally, 6 glass binders at low loadings of 5 mass% were evaluated as sintering aids for the consolidation process. The approach of using the organic Group IV additives can be used to produce large quantities of sodalite at room temperature and shows promise overmore » a method where colloidal silica is used as the silica source. However, the small particle sizes inhibited densification during pressure-less sintering.« less

Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; G.L. Fredrickson; G.G. Galbreth

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.« less

Pyroprocessing of fast flux test facility nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)« less

U.sup.+4 generation in HTER

DOEpatents

Miller, William E [Naperville, IL; Gay, Eddie C [Park Forest, IL; Tomczuk, Zygmunt [Homer Glen, IL

2006-03-14

A improved device and process for recycling spent nuclear fuels, in particular uranium metal, that facilitates the refinement and recovery of uranium metal from spent metallic nuclear fuels. The electrorefiner device comprises two anodes in predetermined spatial relation to a cathode. The anodese have separate current and voltage controls. A much higher voltage than normal for the electrorefining process is applied to the second anode, thereby facilitating oxidization of uranium (III), U.sup.+, to uranium (IV), U.sup.+4. The current path from the second anode to the cathode is physically shorter than the similar current path from the second anode to the spent nuclear fuel contained in a first anode shaped as a basket. The resulting U.sup.+4 oxidizes and solubilizes rough uranium deposited on the surface of the cathode. A softer uranium metal surface is left on the cathode and is more readily removed by a scraper.

Use of Thermodynamic Modeling for Selection of Electrolyte for Electrorefining of Magnesium from Aluminum Alloy Melts

NASA Astrophysics Data System (ADS)

Gesing, Adam J.; Das, Subodh K.

2017-02-01

With United States Department of Energy Advanced Research Project Agency funding, experimental proof-of-concept was demonstrated for RE-12TM electrorefining process of extraction of desired amount of Mg from recycled scrap secondary Al molten alloys. The key enabling technology for this process was the selection of the suitable electrolyte composition and operating temperature. The selection was made using the FactSage thermodynamic modeling software and the light metal, molten salt, and oxide thermodynamic databases. Modeling allowed prediction of the chemical equilibria, impurity contents in both anode and cathode products, and in the electrolyte. FactSage also provided data on the physical properties of the electrolyte and the molten metal phases including electrical conductivity and density of the molten phases. Further modeling permitted selection of electrode and cell construction materials chemically compatible with the combination of molten metals and the electrolyte.

Magnesium Recycling of Partially Oxidized, Mixed Magnesium-Aluminum Scrap through Combined Refining and Solid Oxide Membrane Electrolysis Processes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xiaofei Guan; Peter A. Zink; Uday B. Pal

2012-01-01

Pure magnesium (Mg) is recycled from 19g of partially oxidized 50.5wt.% Mg-Aluminum (Al) alloy. During the refining process, potentiodynamic scans (PDS) were performed to determine the electrorefining potential for magnesium. The PDS show that the electrorefining potential increases over time as the magnesium content inside the Mg-Al scrap decreases. Up to 100% percent of magnesium is refined from the Mg-Al scrap by a novel refining process of dissolving magnesium and its oxide into a flux followed by vapor phase removal of dissolved magnesium and subsequently condensing the magnesium vapor. The solid oxide membrane (SOM) electrolysis process is employed in themore » refining system to enable additional recycling of magnesium from magnesium oxide (MgO) in the partially oxidized Mg-Al scrap. The combination of the refining and SOM processes yields 7.4g of pure magnesium.« less

Criticality safety strategy and analysis summary for the fuel cycle facility electrorefiner at Argonne National Laboratory West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, R.D.; Benedict, R.W.; Lell, R.M.

1996-05-01

As part of the termination activities of Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory (ANL) West, the spent metallic fuel from EBR-II will be treated in the fuel cycle facility (FCF). A key component of the spent-fuel treatment process in the FCF is the electrorefiner (ER) in which the actinide metals are separated from the active metal fission products and the reactive bond sodium. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt, and refined uranium or uranium/plutonium products are deposited at cathodes. The criticality safety strategy and analysis for the ANLmore » West FCF ER is summarized. The FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. To show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOEs) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOEs, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that will verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.« less

Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, R.D.; Benedict, R.W.; Lell, R.M.

1993-09-01

The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutoniummore » products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.« less

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

Magnesium Recycling of Partially Oxidized, Mixed Magnesium-Aluminum Scrap Through Combined Refining and Solid Oxide Membrane (SOM) Electrolysis Processes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guan, Xiaofei; Zink, Peter; Pal, Uday

2012-03-11

Pure magnesium (Mg) is recycled from 19g of partially oxidized 50.5wt.%Mg-Aluminum (Al) alloy. During the refining process, potentiodynamic scans (PDS) were performed to determine the electrorefining potential for magnesium. The PDS show that the electrorefining potential increases over time as the Mg content inside the Mg-Al scrap decreases. Up to 100% percent of magnesium is refined from the Mg-Al scrap by a novel refining process of dissolving magnesium and its oxide into a flux followed by vapor phase removal of dissolved magnesium and subsequently condensing the magnesium vapors in a separate condenser. The solid oxide membrane (SOM) electrolysis process ismore » employed in the refining system to enable additional recycling of magnesium from magnesium oxide (MgO) in the partially oxidized Mg-Al scrap. The combination of the refining and SOM processes yields 7.4g of pure magnesium; could not collect and weigh all of the magnesium recovered.« less

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thornbury, Matthew

Electrorefining (ER) is a major part of efforts at the Y-12 National Security Complex to revolutionize the reprocessing and purification of enriched uranium (EU). Successful implementation of ER could drastically reduce the operational costs and footprint, hazardous materials use, and waste generation.

Electrochemical separation of uranium in the molten system LiF-NaF-KF-UF4

NASA Astrophysics Data System (ADS)

Korenko, M.; Straka, M.; Szatmáry, L.; Ambrová, M.; Uhlíř, J.

2013-09-01

This article is focused on the electrochemical investigation (cyclic voltammetry and related studies) of possible reduction of U4+ ions to metal uranium in the molten system LiF-NaF-KF(eut.)-UF4 that can provide basis for the electrochemical extraction of uranium from molten salts. Two-step reduction mechanism for U4+ ions involving one electron exchange in soluble/soluble U4+/U3+ system and three electrons exchange in the second step were found on the nickel working electrode. Both steps were found to be reversible and diffusion controlled. Based on cyclic voltammetry, the diffusion coefficients of uranium ions at 530 °C were found to be D(U4+) = 1.64 × 10-5 cm2 s-1 and D(U3+) 1.76 × 10-5 cm2 s-1. Usage of the nickel spiral electrode for electrorefining of uranium showed fairly good feasibility of its extraction. However some oxidant present during the process of electrorefining caused that the solid deposits contained different uranium species such as UF3, UO2 and K3UO2F5.

Multi-Physics Modeling of Molten Salt Transport in Solid Oxide Membrane (SOM) Electrolysis and Recycling of Magnesium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powell, Adam; Pati, Soobhankar

2012-03-11

Solid Oxide Membrane (SOM) Electrolysis is a new energy-efficient zero-emissions process for producing high-purity magnesium and high-purity oxygen directly from industrial-grade MgO. SOM Recycling combines SOM electrolysis with electrorefining, continuously and efficiently producing high-purity magnesium from low-purity partially oxidized scrap. In both processes, electrolysis and/or electrorefining take place in the crucible, where raw material is continuously fed into the molten salt electrolyte, producing magnesium vapor at the cathode and oxygen at the inert anode inside the SOM. This paper describes a three-dimensional multi-physics finite-element model of ionic current, fluid flow driven by argon bubbling and thermal buoyancy, and heat andmore » mass transport in the crucible. The model predicts the effects of stirring on the anode boundary layer and its time scale of formation, and the effect of natural convection at the outer wall. MOxST has developed this model as a tool for scale-up design of these closely-related processes.« less

Potentiometric Sensor for Real-Time Remote Surveillance of Actinides in Molten Salts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natalie J. Gese; Jan-Fong Jue; Brenda E. Serrano

2012-07-01

A potentiometric sensor is being developed at the Idaho National Laboratory for real-time remote surveillance of actinides during electrorefining of spent nuclear fuel. During electrorefining, fuel in metallic form is oxidized at the anode while refined uranium metal is reduced at the cathode in a high temperature electrochemical cell containing LiCl-KCl-UCl3 electrolyte. Actinides present in the fuel chemically react with UCl3 and form stable metal chlorides that accumulate in the electrolyte. This sensor will be used for process control and safeguarding of activities in the electrorefiner by monitoring the concentrations of actinides in the electrolyte. The work presented focuses onmore » developing a solid-state cation conducting ceramic sensor for detecting varying concentrations of trivalent actinide metal cations in eutectic LiCl-KCl molten salt. To understand the basic mechanisms for actinide sensor applications in molten salts, gadolinium was used as a surrogate for actinides. The ß?-Al2O3 was selected as the solid-state electrolyte for sensor fabrication based on cationic conductivity and other factors. In the present work Gd3+-ß?-Al2O3 was prepared by ion exchange reactions between trivalent Gd3+ from GdCl3 and K+-, Na+-, and Sr2+-ß?-Al2O3 precursors. Scanning electron microscopy (SEM) was used for characterization of Gd3+-ß?-Al2O3 samples. Microfocus X-ray Diffraction (µ-XRD) was used in conjunction with SEM energy dispersive X-ray spectroscopy (EDS) to identify phase content and elemental composition. The Gd3+-ß?-Al2O3 materials were tested for mechanical and chemical stability by exposing them to molten LiCl-KCl based salts. The effect of annealing on the exchanged material was studied to determine improvements in material integrity post ion exchange. The stability of the ß?-Al2O3 phase after annealing was verified by µ-XRD. Preliminary sensor tests with different assembly designs will also be presented.« less

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.M. Frank

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomicmore » Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.« less

FY-16 Technology Gap Study Technical Report: Analysis of Undissolved Anode Materials of Mark-IV Electrorefiner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoo, Tae-Sic; Vaden, DeeEarl; Westphal, Brian Robert

2016-01-01

The Experimental Breeder Reactor II (EBR-II) is a sodium cooled fast reactor developed at Argonne National Laboratory (ANL). The used fuels from the EBR-II are currently being treated in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL). The Mark IV (Mk-IV) electrorefiner (ER) is a unit process in the FCF, which is primarily assigned to treating the used driver fuels. The stainless steel anode baskets hold the chopped spent driver fuel segments. During electrorefining, the anode baskets are immersed into the electrolyte and the used fuel is dissolved electrochemically. Perforated sides and bottoms allow the flow ofmore » the electrolyte into and out of the anode baskets. The steel cathode is also immersed into the electrolyte and collects the reduced products. The active metal contents in the used fuel (e.g., Cs, Sr, lanthanides, Pu, etc.) reacts with uranium cations in the electrolyte and progressively reports to the electrolyte. Noble metals are mostly retained in the cladding hulls. Varying quantities of zirconium are retained in the cladding hulls depending on the operational conditions of the Mk-IV ER. The undissolved anode materials are removed from the anode baskets and stored for subsequent metal waste form processing. These undissolved materials typically include undissolved fuels, stainless steel cladding, and adhering electrolyte. A couple of hulls are retrieved for chemical analysis and used for estimating the composition of the entire undissolved anode materials. The mass balance attempt based on this practice of estimating the undissolved anode materials has been a challenge due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Responding to the prescribed challenge, this report investigates chemical analysis data as a whole and finds noticeable trends in the compositions of undissolved anode material samples with respect to the mass of the whole undissolved anode materials. Based upon this discovery, an empirical model is proposed.« less

Separation behaviors of actinides from rare-earths in molten salt electrorefining using saturated liquid cadmium cathode

NASA Astrophysics Data System (ADS)

Kato, Tetsuya; Inoue, Tadashi; Iwai, Takashi; Arai, Yasuo

2006-10-01

Electrorefining in the molten LiCl-KCl eutectic salt containing actinide (An) and rare-earth (RE) elements was conducted to recover An elements up to 10 wt% into liquid cadmium (Cd) cathode, which is much higher than the solubility of the An elements in liquid Cd at the experimental temperature of 773 K. In the saturated Cd cathode, the An and RE elements were recovered forming a PuCd 11 type compound, MCd 11 (M = An and RE elements). The separation factors of element M against Pu defined as [M/Pu in Cd alloy (cathode)]/[M/Pu in molten salt] were calculated for the saturated Cd cathode including MCd 11. The separation factors were 0.011, 0.044, 0.064, and 0.064 for La, Ce, Pr, and Nd, respectively. These values were a little differed from 0.014, 0.038, 0.044, and 0.043 for the equilibrium unsaturated liquid Cd, respectively. The above slight differences were considered to be caused by the solid phase formation in the saturated Cd cathode and the electrochemical transfer of the An and RE elements in the molten salt.

Extraterrestrial materials processing and construction. [space industrialization

NASA Technical Reports Server (NTRS)

Criswell, D. R.; Waldron, R. D.; Mckenzie, J. D.

1980-01-01

Three different chemical processing schemes were identified for separating lunar soils into the major oxides and elements. Feedstock production for space industry; an HF acid leach process; electrorefining processes for lunar free metal and metal derived from chemical processing of lunar soils; production and use of silanes and spectrally selective materials; glass, ceramics, and electrochemistry workshops; and an econometric model of bootstrapping space industry are discussed.

Method For Processing Spent (Trn,Zr)N Fuel

DOEpatents

Miller, William E.; Richmann, Michael K.

2004-07-27

A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

Developing a Signature Based Safeguards Approach for the Electrorefiner and Salt Cleanup Unit Operations in Pyroprocessing Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Chantell Lynne-Marie

Traditional nuclear materials accounting does not work well for safeguards when applied to pyroprocessing. Alternate methods such as Signature Based Safeguards (SBS) are being investigated. The goal of SBS is real-time/near-real-time detection of anomalous events in the pyroprocessing facility as they could indicate loss of special nuclear material. In high-throughput reprocessing facilities, metric tons of separated material are processed that must be accounted for. Even with very low uncertainties of accountancy measurements (<0.1%) the uncertainty of the material balances is still greater than the desired level. Novel contributions of this work are as follows: (1) significant enhancement of SBS developmentmore » for the salt cleanup process by creating a new gas sparging process model, selecting sensors to monitor normal operation, identifying safeguards-significant off-normal scenarios, and simulating those off-normal events and generating sensor output; (2) further enhancement of SBS development for the electrorefiner by simulating off-normal events caused by changes in salt concentration and identifying which conditions lead to Pu and Cm not tracking throughout the rest of the system; and (3) new contribution in applying statistical techniques to analyze the signatures gained from these two models to help draw real-time conclusions on anomalous events.« less

Safeguards in Pyroprocessing: an Integrated Model Development and Measurement Data Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Jinsuo

Pyroprocessing is an electrochemical method based on the molten salt electrolyte, mainly the LiCl-KCl eutectic molten salt, to recycle the used nuclear fuel. For a conceptual design of commercial pyroprocessing facility, tons of special nuclear materials, namely U and Pu, may be involved, which could be used for non-peaceful purposes if they are diverted. Effective safeguards approaches have to be developed prior to the development and construction of a pyroprocessing facility. Present research focused on two main objectives, namely calculating the properties of nuclear species in LiCl-KCl molten salt and developing integrated model to safeguard a pyroprocessing facility. Understanding themore » characteristics of special nuclear materials in LiCl-KCl eutectic salt is extremely important to understand their behaviors in an electrorefiner. The model development for the separation processes in the pyroprocessing, including electrorefining, actinide drawdown, and rare earth drawdown benefits the understanding of material transport and separation performance of these processes under various conditions. The output signals, such as potential, current, and species concentration contribute to the material balance closure and provide safeguards signatures to detect the scenarios of diversion. U and Pu are the two main elements concerned in this study due to our interest in safeguards.« less

Recycling of Magnesium Alloy Employing Refining and Solid Oxide Membrane (SOM) Electrolysis

NASA Astrophysics Data System (ADS)

Guan, Xiaofei; Zink, Peter A.; Pal, Uday B.; Powell, Adam C.

2013-04-01

Pure magnesium was recycled from partially oxidized 50.5 wt pct Mg-Al scrap alloy and AZ91 Mg alloy (9 wt pct Al, 1 wt pct Zn). Refining experiments were performed using a eutectic mixture of MgF2-CaF2 molten salt (flux). During the experiments, potentiodynamic scans were performed to determine the electrorefining potentials for magnesium dissolution and magnesium bubble nucleation in the flux. The measured electrorefining potential for magnesium bubble nucleation increased over time as the magnesium content inside the magnesium alloy decreased. Potentiostatic holds and electrochemical impedance spectroscopy were employed to measure the electronic and ionic resistances of the flux. The electronic resistivity of the flux varied inversely with the magnesium solubility. Up to 100 pct of the magnesium was refined from the Mg-Al scrap alloy by dissolving magnesium and its oxide into the flux followed by argon-assisted evaporation of dissolved magnesium and subsequently condensing the magnesium vapor. Solid oxide membrane electrolysis was also employed in the system to enable additional magnesium recovery from magnesium oxide in the partially oxidized Mg-Al scrap. In an experiment employing AZ91 Mg alloy, only the refining step was carried out. The calculated refining yield of magnesium from the AZ91 alloy was near 100 pct.

Supported liquid membrane electrochemical separators

DOEpatents

Pemsler, J. Paul; Dempsey, Michael D.

1986-01-01

Supported liquid membrane separators improve the flexibility, efficiency and service life of electrochemical cells for a variety of applications. In the field of electrochemical storage, an alkaline secondary battery with improved service life is described in which a supported liquid membrane is interposed between the positive and negative electrodes. The supported liquid membranes of this invention can be used in energy production and storage systems, electrosynthesis systems, and in systems for the electrowinning and electrorefining of metals.

Fuel conditioning facility electrorefiner start-up results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Mariani, R.D.; Vaden, D.

1996-05-01

At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete.

Investigation of residual anode material after electrorefining uranium in molten chloride salt

NASA Astrophysics Data System (ADS)

Rose, M. A.; Williamson, M. A.; Willit, J.

2015-12-01

A buildup of material at uranium anodes during uranium electrorefining in molten chloride salts has been observed. Potentiodynamic testing has been conducted using a three electrode cell, with a uranium working electrode in both LiCl/KCl eutectic and LiCl each containing ∼5 mol% UCl3. The anodic current response was observed at 50° intervals between 450 °C and 650 °C in the eutectic salt. These tests revealed a buildup of material at the anode in LiCl/KCl salt, which was sampled at room temperature, and analyzed using ICP-MS, XRD and SEM techniques. Examination of the analytical data, current response curves and published phase diagrams has established that as the uranium anode dissolves, the U3+ ion concentration in the diffusion layer surrounding the electrode rises precipitously to levels, which may at low temperatures exceed the solubility limit for UCl3 or in the case of the eutectic salt for K2UCl5. The reduction in current response observed at low temperature in eutectic salt is eliminated at 650 °C, where K2UCl5 is absent due to its congruent melting and only simple concentration polarization effects are seen. In LiCl similar concentration effects are seen though significantly longer time at applied potential is required to effect a reduction in the current response as compared to the eutectic salt.

Corrosion Behavior of Yttria-Stabilized Zirconia-Coated 9Cr-1Mo Steel in Molten UCl3-LiCl-KCl Salt

NASA Astrophysics Data System (ADS)

Jagadeeswara Rao, Ch.; Venkatesh, P.; Prabhakara Reddy, B.; Ningshen, S.; Mallika, C.; Kamachi Mudali, U.

2017-02-01

For the electrorefining step in the pyrochemical reprocessing of spent metallic fuels of future sodium cooled fast breeder reactors, 9Cr-1Mo steel has been proposed as the container material. The electrorefining process is carried out using 5-6 wt.% UCl3 in LiCl-KCl molten salt as the electrolyte at 500 °C under argon atmosphere. In the present study, to protect the container vessel from hot corrosion by the molten salt, 8-9% yttria-stabilized zirconia (YSZ) ceramic coating was deposited on 9Cr-1Mo steel by atmospheric plasma spray process. The hot corrosion behavior of YSZ-coated 9Cr-1Mo steel specimen was investigated in molten UCl3-LiCl-KCl salt at 600 °C for 100-, 500-, 1000- and 2000-h duration. The results revealed that the weight change in the YSZ-coated specimen was insignificant even after exposure to molten salt for 2000 h, and delamination of coating did not occur. SEM examination showed the lamellar morphology of the YSZ coating after the corrosion test with occluded molten salt. The XRD analysis confirmed the presence of tetragonal and cubic phases of ZrO2, without any phase change. Formation of UO2 in some regions of the samples was evident from XRD results.

Membrane Purification Cell for Aluminum Recycling

DOE Office of Scientific and Technical Information (OSTI.GOV)

David DeYoung; James Wiswall; Cong Wang

2011-11-29

Recycling mixed aluminum scrap usually requires adding primary aluminum to the scrap stream as a diluent to reduce the concentration of non-aluminum constituents used in aluminum alloys. Since primary aluminum production requires approximately 10 times more energy than melting scrap, the bulk of the energy and carbon dioxide emissions for recycling are associated with using primary aluminum as a diluent. Eliminating the need for using primary aluminum as a diluent would dramatically reduce energy requirements, decrease carbon dioxide emissions, and increase scrap utilization in recycling. Electrorefining can be used to extract pure aluminum from mixed scrap. Some example applications includemore » producing primary grade aluminum from specific scrap streams such as consumer packaging and mixed alloy saw chips, and recycling multi-alloy products such as brazing sheet. Electrorefining can also be used to extract valuable alloying elements such as Li from Al-Li mixed scrap. This project was aimed at developing an electrorefining process for purifying aluminum to reduce energy consumption and emissions by 75% compared to conventional technology. An electrolytic molten aluminum purification process, utilizing a horizontal membrane cell anode, was designed, constructed, operated and validated. The electrorefining technology could also be used to produce ultra-high purity aluminum for advanced materials applications. The technical objectives for this project were to: - Validate the membrane cell concept with a lab-scale electrorefining cell; - Determine if previously identified voltage increase issue for chloride electrolytes holds for a fluoride-based electrolyte system; - Assess the probability that voltage change issues can be solved; and - Conduct a market and economic analysis to assess commercial feasibility. The process was tested using three different binary alloy compositions (Al-2.0 wt.% Cu, Al-4.7 wt.% Si, Al-0.6 wt.% Fe) and a brazing sheet scrap composition (Al-2.8 wt.% Si-0.7 wt.% Fe-0.8 wt.% Mn),. Purification factors (defined as the initial impurity concentration divided by the final impurity concentration) of greater than 20 were achieved for silicon, iron, copper, and manganese. Cell performance was measured using its current and voltage characteristics and composition analysis of the anode, cathode, and electrolytes. The various cells were autopsied as part of the study. Three electrolyte systems tested were: LiCl-10 wt. % AlCl3, LiCl-10 wt. % AlCl3-5 wt.% AlF3 and LiF-10 wt.% AlF3. An extended four-day run with the LiCl-10 wt.% AlCl3-5 wt.% AlF3 electrolyte system was stable for the entire duration of the experiment, running at energy requirements about one third of the Hoopes and the conventional Hall-Heroult process. Three different anode membranes were investigated with respect to their purification performance and survivability: a woven graphite cloth with 0.05 cm nominal thickness & > 90 % porosity, a drilled rigid membrane with nominal porosity of 33%, and another drilled rigid graphite membrane with increased thickness. The latter rigid drilled graphite was selected as the most promising membrane design. The economic viability of the membrane cell to purify scrap is sensitive to primary & scrap aluminum prices, and the cost of electricity. In particular, it is sensitive to the differential between scrap and primary aluminum price which is highly variable and dependent on the scrap source. In order to be economically viable, any scrap post-processing technology in the U.S. market must have a total operating cost well below the scrap price differential of $0.20-$0.40 per lb to the London Metal Exchange (LME), a margin of 65%-85% of the LME price. The cost to operate the membrane cell is estimated to be < $0.24/lb of purified aluminum. The energy cost is estimated to be $0.05/lb of purified aluminum with the remaining costs being repair and maintenance, electrolyte, labor, taxes and depreciation. The bench-scale work on membrane purification cell process has demonstrated technological advantages and substantial energy and investment savings against other electrolytic processes. However, in order to realize commercial reality, the following items need to be fully investigated: 1. Further evaluation of a pure fluoride electrolyte. 2. Investigate alternative non conductive, more mechanically robust and chemically inert membrane candidates. 3. Optimized membrane cell design to understand contribution of fluid flow patterns and the mass transfer conditions. 4. Improve current efficiency and total metallic aluminum recovery from the cell. All Tasks and Milestones were completed successfully.« less

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which hasmore » the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.« less

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rechard, Robert P.; Trone, Janis R.; Kalinina, Elena Arkadievna

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a minedmore » repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.« less

Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Jinsuo; Guo, Shaoqiang

Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels.

A phase-field simulation of uranium dendrite growth on the cathode in the electrorefining process

NASA Astrophysics Data System (ADS)

Shibuta, Yasushi; Unoura, Seiji; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Suzuki, Toshio

2011-07-01

The uranium dendrite growth on the cathode during the pyroprocessing of uranium is investigated using a novel phase-field model, in which electrodeposition of uranium and zirconium from the molten-salt is taken into account. The threshold concentration of zirconium in the molten salt demarcating the dendritic and planar growth is then estimated as a function of the current density. Moreover, the growth process of both the dendritic and planar electrodeposits has been demonstrated by way of varying the mobility of the phase field, which consists of the effect of attachment kinetics and diffusion.

The use of magnesium in lightweight lithium-ion battery packs

NASA Astrophysics Data System (ADS)

Neelameggham, Neale R.

2009-04-01

The analysis of recently announced battery packs for plug-in hybrid electric vehicles (PHEV) shows that the design of the series-parallel combinations is being over-complicated. The proven energy densities of lithium-ion cells from about 200 Wh/kg are being reduced to 90 Wh/kg. The majority of the weight increase seems to be for thermal management. Simpler battery pack designs based on electro-refining pot rooms using self-contained rectangular lithium-ion cells with air cooling inside of die-cast magnesium cell tanks would help avoid hauling dead weight in PHEV by providing considerable weight reduction.

Electrorefiner

DOEpatents

Miller, W.E.; Tomczuk, Z.

1995-08-22

An apparatus is disclosed capable of functioning as a solid cathode and for removing crystalline structure from the upper surface of a liquid cathode, includes a metallic support vertically disposed with respect to an electrically insulating container capable of holding a liquid metal cathode. A piston of electrically insulating material mounted on the drive tube, surrounding the current lead, for vertical and rotational movement with respect thereto including a downwardly extending collar portion surrounding the metallic current lead. At least one portion of the piston remote from the metallic current lead being removed. Mechanism for lowering the piston to the surface of the liquid cathode and raising the piston from the surface along with mechanism for rotating the piston around its longitudinal axis. 5 figs.

Electrorefiner

DOEpatents

Miller, William E.; Tomczuk, Zygmunt

1995-01-01

An apparatus capable of functioning as a solid cathode and for removing crystalline structure from the upper surface of a liquid cathode, includes a metallic support vertically disposed with respect to an electrically insulating container capable of holding a liquid metal cathode. A piston of electrically insulating material mounted on the drive tube, surrounding the current lead, for vertical and rotational movement with respect thereto including a downwardly extending collar portion surrounding the metallic current lead. At least one portion of the piston remote from the metallic current lead being removed. Mechanism for lowering the piston to the surface of the liquid cathode and raising the piston from the surface along with mechanism for rotating the piston around its longitudinal axis.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; Smith, Paul Herrick; Jarvinen, Gordon D.

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that themore » following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl 3 by the precipitation of PuPO 4.« less

Development and Optimization of Voltammetric Methods for Real Time Analysis of Electrorefiner Salt with High Concentrations of Actinides and Fission Products

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.; Phongikaroon, Supathorn; Zhang, Jinsuo

This project addresses the problem of achieving accurate material control and accountability (MC&A) around pyroprocessing electrorefiner systems. Spent nuclear fuel pyroprocessing poses a unique challenge with respect to reprocessing technology in that the fuel is never fully dissolved in the process fluid. In this case, the process fluid is molten, anhydrous LiCl-KCl salt. Therefore, there is no traditional input accountability tank. However, electrorefiners (ER) accumulate very large quantities of fissile nuclear material (including plutonium) and should be well safeguarded in a commercial facility. Idaho National Laboratory (INL) currently operates a pyroprocessing facility for treatment of spent fuel from Experimental Breedermore » Reactor-II with two such ER systems. INL implements MC&A via a mass tracking model in combination with periodic sampling of the salt and other materials followed by destructive analysis. This approach is projected to be insufficient to meet international safeguards timeliness requirements. A real time or near real time monitoring method is, thus, direly needed to support commercialization of pyroprocessing. A variety of approaches to achieving real time monitoring for ER salt have been proposed and studied to date—including a potentiometric actinide sensor for concentration measurements, a double bubbler for salt depth and density measurements, and laser induced breakdown spectroscopy (LIBS) for concentration measurements. While each of these methods shows some promise, each also involves substantial technical complexity that may ultimately limit their implementation. Yet another alternative is voltammetry—a very simple method in theory that has previously been tested for this application to a limited extent. The equipment for a voltammetry system consists of off-the-shelf components (three electrodes and a potentiostat), which results in substantial benefits relative to cost and robustness. Based on prior knowledge of electrochemical reduction potentials for each of the species of interest, voltammetry can be used to quantify concentrations of a variety of elemental species—including uranium, plutonium, minor actinides, and rare earths. Various methods have been tested by other researchers to date—including cyclic voltammetry, square wave voltammetry, normal pulse voltammetry, etc. In most cases, it has been observed that there is a very limited concentration range for which the output can be readily correlated with concentration in the salt. Furthermore, testing to date has been limited to simple ternary salts with only a single element being quantified. While incomplete for application to MC&A for pyroprocessing, these results lead us to believe that voltammetry can be optimized based on salt properties and fundamental electrochemical rate processes to yield a highly accurate and robust method. This project is divided into four tasks jointly executed by three university research groups. This includes experimental measurement of key physical data on the systems of interest, development of a predictive voltammetry model, experimental validation of the voltammetry model, and design/verification of an optimized measurement method. This project supports the goals of the US-ROK Joint Fuel Cycle Study in addition to the NA-24 Office of the National Nuclear Security Agency and the International Atomic Energy Agency (IAEA).« less

Morphology of uranium electrodeposits on cathode in electrorefining process: A phase-field simulation

NASA Astrophysics Data System (ADS)

Shibuta, Yasushi; Sato, Takumi; Suzuki, Toshio; Ohta, Hirokazu; Kurata, Masaki

2013-05-01

Morphology of uranium electrodeposits on cathode with respect to applied voltage, zirconium concentration in the molten salt and the size of primary deposit during pyroprocessing is systematically investigated by the phase-field simulation. It is found that there is a threshold zirconium concentration in the molten salt demarcating planar and cellular/needle-like electrodeposits, which agrees with experimental results. In addition, the effect of size of primary deposits on the morphology of electrodeposits is examined. It is then confirmed that cellular/needle-like electrodeposits are formed from large primary deposits at all applied voltages considered, whereas both the planar and cellular/needle-like electrodeposits are formed from the primary deposits of 10 μm and less.

Electrochemistry of uranium in molten LiF-CaF2

NASA Astrophysics Data System (ADS)

Nourry, C.; Souček, P.; Massot, L.; Malmbeck, R.; Chamelot, P.; Glatz, J.-P.

2012-11-01

This article is focused on the electrochemical behaviour of U ions in molten LiF-CaF2 (79-21 wt.%) eutectic. On a W electrode, U(III) is reduced in one step to U metal and U(III) can be also oxidised to U(IV). Both systems were studied by cyclic and square wave voltammetry. Reversibility of both systems for both techniques was verified and number of exchanged electrons was determined, as well as diffusion coefficients for U(III) and U(IV). The results are in a good agreement with previous studies. On a Ni electrode, the depolarisation effect due to intermetallic compounds formation was observed. Electrorefining of U metal in a melt containing U and Gd ions was carried out using a reactive Ni electrode with promising results.

Magnesium Electrorefining in Non-Aqueous Electrolyte at Room Temperature

NASA Astrophysics Data System (ADS)

Kwon, Kyungjung; Park, Jesik; Kusumah, Priyandi; Dilasari, Bonita; Kim, Hansu; Lee, Churl Kyoung

Magnesium, of which application is often limited by its poor corrosion resistance, is more vulnerable to corrosion with existence of metal impurities such as Fe. Therefore, for the refining and recycling of magnesium, high temperature electrolysis using molten salts has been frequently adopted. In this report, the purification of magnesium scrap by electrolysis at room temperature is investigated with non-aqueous electrolytes. An aprotic solvent of tetrahydrofuran (THF) was used as a solvent of the electrolyte. Magnesium scrap was used as anode materials and ethyl magnesium bromide (EtMgBr) was dissolved in THF for magnesium source. The purified magnesium can be uniformly electrodeposited on copper electrode under potentiostatic conditions. The deposits were confirmed by scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS) analysis.

Ceramic waste form production and development at ANL-West.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Battisti, T. J.; Goff, K. M.; Bateman, K. J.

2002-08-21

Argonne National Laboratory has developed a method to stabilize spent electrolyte salt discarded from electrorefiners (ER) used to treat spent nuclear fuel. The salt is stabilized in a ceramic using a pressureless consolidation technique. The starting material is zeolite 4A which is used as the host for the fission product and actinide rich salt. Glass frit is added to the salt loaded zeolite before processing to act as a binder. The zeolite 4A is converted to sodalite during processing via pressureless consolidation. This process differs from one used in the past that employed a hot isostatic press. Ceramic is createdmore » at 925 C and atmospheric pressure instead of the high pressures used in hot isostatic pressing. Process flow sheets, off-gas test results, processing equipment, and leech test results are presented.« less

Pyrochemical process for extracting plutonium from an electrolyte salt

DOEpatents

Mullins, L.J.; Christensen, D.C.

1982-09-20

A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium for electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

Preparation and Characterization of a Master Blend of Plutonium Oxide for the 3013 Large Scale Shelf-Life Surveillance Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gillispie, Obie William; Worl, Laura Ann; Veirs, Douglas Kirk

A mixture of chlorine-containing, impure plutonium oxides has been produced and has been given the name Master Blend. This large quantity of well-characterized chlorinecontaining material is available for use in the Integrated Surveillance and Monitoring Program for shelf-life experiments. It is intended to be representative of materials packaged to meet DOE-STD-3013.1 The Master Blend contains a mixture of items produced in Los Alamos National Laboratory’s (LANL) electro-refining pyrochemical process in the late 1990s. Twenty items were crushed and sieved, calcined to 800ºC for four hours, and blended multiple times. This process resulted in four batches of Master Blend. Calorimetry andmore » density data on material from the four batches indicate homogeneity.« less

Pyrochemical process for extracting plutonium from an electrolyte salt

DOEpatents

Mullins, Lawrence J.; Christensen, Dana C.

1984-01-01

A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium from electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. D. Herrmann; L. A. Wurth; N. J. Gese

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimentalmore » study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.« less

Phosphates behaviours in conversion of FP chlorides

NASA Astrophysics Data System (ADS)

Amamoto, I.; Kofuji, H.; Myochin, M.; Takasaki, Y.; Terai, T.

2009-06-01

The spent electrolyte of the pyroprocessing by metal electrorefining method should be considered for recycling after removal of fission products (FP) such as, alkali metals (AL), alkaline earth metals (ALE), and/or rare earth elements (REE), to reduce the volume of high-level radioactive waste. Among the various methods suggested for this purpose is precipitation by converting FP from chlorides to phosphates. Authors have been carrying out the theoretical analysis and experiment showing the behaviours of phosphate precipitates so as to estimate the feasibility of this method. From acquired results, it was found that AL except lithium and ALE are unlikely to form phosphate precipitates. However their conversion behaviours including REE were compatible with the theoretical analysis; in the case of LaPO 4 as one of the REE precipitates, submicron-size particles could be observed while that of Li 3PO 4 was larger; the precipitates were apt to grow larger at higher temperature; etc.

Update on Recovering Lead From Scrap Batteries

NASA Astrophysics Data System (ADS)

Cole, E. R.; Lee, A. Y.; Paulson, D. L.

1985-02-01

Previous work at the Bureau of Mines Rolla Research Center, U.S. Department of the Interior, resulted in successful development of a bench-scale, combination electrorefining-electrowinning method for recycling lead from scrap batteries by using waste fluosilicic acid (H2SiF6) as electrolyte.1,2 This paper describes larger scale experiments. Prior attempts to electrowin lead failed because large quantities of insoluble lead dioxide were deposited on the anodes at the expense of lead deposition on the cathodes. A major breakthrough was achieved with the discovery that lead dioxide formation at the anodes is prevented by adding a small amount of phosphorus to the electrolyte. The amount of PbO2 formed on the anodes during lead electrowinning was less than 1% of the total lead deposited on the cathodes. This work recently won the prestigious IR·100 award as one of the 100 most significant technological advances of 1984.

A finite difference model used to predict the consolidation of a ceramic waste form produced from the electrometallurgical treatment of spent nuclear fuel.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bateman, K. J.; Capson, D. D.

2004-03-29

Argonne National Laboratory (ANL) has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in the electrometallurgical treatment of spent Experimental Breeder Reactor-II (EBR-II) fuel. The ceramic waste process culminates with an elevated temperature operation. The processing conditions used by the furnace, for demonstration scale and production scale operations, are to be developed at Argonne National Laboratory-West (ANL-West). To assist in selecting the processing conditions of the furnace and to reduce the number of costly experiments, a finitemore » difference model was developed to predict the consolidation of the ceramic waste. The model accurately predicted the heating as well as the bulk density of the ceramic waste form. The methodology used to develop the computer model and a comparison of the analysis to experimental data is presented.« less

Projected Salt Waste Production from a Commercial Pyroprocessing Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.

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 separatingmore » 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.« less

Heteroepitaxial diamond growth on 4H-SiC using microwave plasma chemical vapor deposition.

PubMed

Moore, Eric; Jarrell, Joshua; Cao, Lei

2017-09-01

Deposition of heteroepitaxial diamond via microwave chemical vapor deposition has been performed on a 4H-SiC substrate using bias enhanced nucleation followed by a growth step. In future work, the diamond film will serve as a protective layer for an alpha particle sensor designed to function in an electrorefiner during pyroprocessing of spent fuel. The diamond deposition on the 4H-SiC substrate was carried out using a methane-hydrogen gas mixture with varying gas flow rates. The nucleation step was conducted for 30 minutes and provided sufficient nucleation sites to grow a diamond film on various locations on the substrate. The resulting diamond film was characterized using Raman spectroscopy exhibiting the strong Raman peak at 1332 cm -1 . Scanning electron microscopy was used to observe the surface morphology and the average grain size of the diamond film was observed to be on the order of ∼2-3 μm.

Temperature effect on laser-induced breakdown spectroscopy spectra of molten and solid salts

NASA Astrophysics Data System (ADS)

Hanson, Cynthia; Phongikaroon, Supathorn; Scott, Jill R.

2014-07-01

Laser-induced breakdown spectroscopy (LIBS) has been investigated as a potential analytical tool to improve operations and safeguards for electrorefiners, such as those used in processing spent nuclear fuel. This study set out to better understand the effect of sample temperature and physical state on LIBS spectra of molten and solid salts by building calibration curves of cerium and assessing self-absorption, plasma temperature, electron density, and local thermal equilibrium (LTE). Samples were composed of a LiCl-KCl eutectic salt, an internal standard of MnCl2, and varying concentrations of CeCl3 (0.1, 0.3, 0.5, 0.8, and 1.0 wt.% Ce) under different temperatures (773, 723, 673, 623, and 573 K). Analysis of salts in their molten form is preferred as plasma plumes from molten samples experienced less self-absorption, less variability in plasma temperature, and higher clearance of the minimum electron density required for local thermal equilibrium. These differences are attributed to plasma dynamics as a result of phase changes. Spectral reproducibility was also better in the molten state due to sample homogeneity.

Monte Carlo simulations of safeguards neutron counter for oxide reduction process feed material

NASA Astrophysics Data System (ADS)

Seo, Hee; Lee, Chaehun; Oh, Jong-Myeong; An, Su Jung; Ahn, Seong-Kyu; Park, Se-Hwan; Ku, Jeong-Hoe

2016-10-01

One of the options for spent-fuel management in Korea is pyroprocessing whose main process flow is the head-end process followed by oxide reduction, electrorefining, and electrowining. In the present study, a well-type passive neutron coincidence counter, namely, the ACP (Advanced spent fuel Conditioning Process) safeguards neutron counter (ASNC), was redesigned for safeguards of a hot-cell facility related to the oxide reduction process. To this end, first, the isotopic composition, gamma/neutron emission yield and energy spectrum of the feed material ( i.e., the UO2 porous pellet) were calculated using the OrigenARP code. Then, the proper thickness of the gammaray shield was determined, both by irradiation testing at a standard dosimetry laboratory and by MCNP6 simulations using the parameters obtained from the OrigenARP calculation. Finally, the neutron coincidence counter's calibration curve for 100- to 1000-g porous pellets, in consideration of the process batch size, was determined through simulations. Based on these simulation results, the neutron counter currently is under construction. In the near future, it will be installed in a hot cell and tested with spent fuel materials.

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Pierce, David A.; Crum, Jarrod V.

The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change.more » The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.« less

Method for removing metal vapor from gas streams

DOEpatents

Ahluwalia, R.K.; Im, K.H.

1996-04-02

A process for cleaning an inert gas contaminated with a metallic vapor, such as cadmium, involves withdrawing gas containing the metallic contaminant from a gas atmosphere of high purity argon; passing the gas containing the metallic contaminant to a mass transfer unit having a plurality of hot gas channels separated by a plurality of coolant gas channels; cooling the contaminated gas as it flows upward through the mass transfer unit to cause contaminated gas vapor to condense on the gas channel walls; regenerating the gas channels of the mass transfer unit; and, returning the cleaned gas to the gas atmosphere of high purity argon. The condensing of the contaminant-containing vapor occurs while suppressing contaminant particulate formation, and is promoted by providing a sufficient amount of surface area in the mass transfer unit to cause the vapor to condense and relieve supersaturation buildup such that contaminant particulates are not formed. Condensation of the contaminant is prevented on supply and return lines in which the contaminant containing gas is withdrawn and returned from and to the electrorefiner and mass transfer unit by heating and insulating the supply and return lines. 13 figs.

Recovery of actinides from actinide-aluminium alloys by chlorination: Part III - Chlorination with HCl(g)

NASA Astrophysics Data System (ADS)

Meier, Roland; Souček, Pavel; Walter, Olaf; Malmbeck, Rikard; Rodrigues, Alcide; Glatz, Jean-Paul; Fanghänel, Thomas

2018-01-01

Two steps of a pyrochemical route for the recovery of actinides from spent metallic nuclear fuel are being investigated at JRC-Karlsruhe. The first step consists in electrorefining the fuel in molten salt medium implying aluminium cathodes. The second step is a chlorination process for the separation of actinides (An) from An-Al alloys formed on the cathodes. The chlorination process, in turn, consists of three steps; the distillation of adhered salt (1), the chlorination of An-Al by HCl/Cl2 under formation of AlCl3 and An chlorides (2), and the subsequent sublimation of AlCl3 (3). In the present work UAl2, UAl3, NpAl2, and PuAl2 were chlorinated with HCl(g) in a temperature range between 300 and 400 °C forming UCl4, NpCl4 or PuCl3 as the major An containing phases, respectively. Thermodynamic calculations were carried out to support the experimental work. The results showed a high chlorination efficiency for all used starting materials and indicated that the sublimation step may not be necessary when using HCl(g).

JOWOG 22/2 - Actinide Chemical Technology (July 9-13, 2012)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, Jay M.; Lopez, Jacquelyn C.; Wayne, David M.

2012-07-05

The Plutonium Science and Manufacturing Directorate provides world-class, safe, secure, and reliable special nuclear material research, process development, technology demonstration, and manufacturing capabilities that support the nation's defense, energy, and environmental needs. We safely and efficiently process plutonium, uranium, and other actinide materials to meet national program requirements, while expanding the scientific and engineering basis of nuclear weapons-based manufacturing, and while producing the next generation of nuclear engineers and scientists. Actinide Process Chemistry (NCO-2) safely and efficiently processes plutonium and other actinide compounds to meet the nation's nuclear defense program needs. All of our processing activities are done in amore » world class and highly regulated nuclear facility. NCO-2's plutonium processing activities consist of direct oxide reduction, metal chlorination, americium extraction, and electrorefining. In addition, NCO-2 uses hydrochloric and nitric acid dissolutions for both plutonium processing and reduction of hazardous components in the waste streams. Finally, NCO-2 is a key team member in the processing of plutonium oxide from disassembled pits and the subsequent stabilization of plutonium oxide for safe and stable long-term storage.« less

Method for removing metal vapor from gas streams

DOEpatents

Ahluwalia, R. K.; Im, K. H.

1996-01-01

A process for cleaning an inert gas contaminated with a metallic vapor, such as cadmium, involves withdrawing gas containing the metallic contaminant from a gas atmosphere of high purity argon; passing the gas containing the metallic contaminant to a mass transfer unit having a plurality of hot gas channels separated by a plurality of coolant gas channels; cooling the contaminated gas as it flows upward through the mass transfer unit to cause contaminated gas vapor to condense on the gas channel walls; regenerating the gas channels of the mass transfer unit; and, returning the cleaned gas to the gas atmosphere of high purity argon. The condensing of the contaminant-containing vapor occurs while suppressing contaminant particulate formation, and is promoted by providing a sufficient amount of surface area in the mass transfer unit to cause the vapor to condense and relieve supersaturation buildup such that contaminant particulates are not formed. Condensation of the contaminant is prevented on supply and return lines in which the contaminant containing gas is withdrawn and returned from and to the electrorefiner and mass transfer unit by heating and insulating the supply and return lines.

DYNAMIC PROPERTIES OF SHOCK LOADED THIN URANIUM FOILS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robbins, D. L.; Kelly, A. M.; Alexander, D. J.

A series of spall experiments has been completed with thin depleted uranium targets, nominally 0.1 mm thick. The first set of uranium spall targets was cut and ground to final thickness from electro-refined, high-purity, cast uranium. The second set was rolled to final thickness from low purity uranium. The impactors for these experiments were laser-launched 0.05-mm thick copper flyers, 3 mm in diameter. Laser energies were varied to yield a range of flyer impact velocities. This resulted in varying degrees of damage to the uranium spall targets, from deformation to complete spall or separation at the higher velocities. Dynamic measurementsmore » of the uranium target free surface velocities were obtained with dual velocity interferometers. Uranium targets were recovered and sectioned after testing. Free surface velocity profiles were similar for the two types of uranium, but spall strengths (estimated from the magnitude of the pull-back signal) are higher for the high-purity cast uranium. Velocity profiles and microstructural evidence of spall from the sectioned uranium targets are presented.« less

Vacuum distillation of a mixture of LiCl-KCl eutectic salts and RE oxidative precipitates and a dechlorination and oxidation of RE oxychlorides.

PubMed

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

2008-12-30

In this study, a vacuum distillation of a mixture of LiCl-KCl eutectic salt and rare-earth oxidative precipitates was performed to separate a pure LiCl-KCl eutectic salt from the mixture. Also, a dechlorination and oxidation of the rare-earth oxychlorides was carried out to stabilize a final waste form. The mixture was distilled under a range of 710-759.5Torr of a reduced pressure at a fixed heating rate of 4 degrees C/min and the LiCl-KCl eutectic salt was completely separated from the mixture. The required time for the salt distillation and the starting temperature for the salt vaporization were lowered with a reduction in the pressure. Dechlorination and oxidation of the rare-earth oxychlorides was completed at a temperature below 1300 degrees C and this was dependent on the partial pressure of O2. The rare-earth oxychlorides (NdOCl/PrOCl) were transformed to oxides (Nd2O3/PrO2) during the dechlorination and oxidation process. These results will be utilized to design a concept for a process for recycling the waste salt from an electrorefining process.

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Pierce, David A.; Crum, Jarrod V.

The work presented here is based off the concept that the rare earth chloride (RECl 3) fission products within the used electrorefiner (ER) salt can be selectively removed as RECl 3 (not yet demonstrated) or precipitated out as a mixture of REOCl and REO x through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl 3s at 10 mass% into a 78%TeO 2-22%PbO glass while also showing that this same mixture of RECl 3s can be oxidized to REOCl at 300 °C and then to REO x by 1200 °C, evolvingmore » Cl 2(g). When the REO x mixture is heated at temperatures >1200 °C, the ratios of REO xs change. The mixture of REO x was then immobilized in a lanthanide borosilicate (LABS) glass at a high loading of 60 mass%. Both the 78%TeO 2-22%PbO glass and LABS glass systems show good chemical durability. In conclusion, the advantages and disadvantages of tellurite and LABS glasses are compared.« less

Performance evaluation of PRIDE UNDA system with pyroprocessing feed material.

PubMed

An, Su Jung; Seo, Hee; Lee, Chaehun; Ahn, Seong-Kyu; Park, Se-Hwan; Ku, Jeong-Hoe

2017-04-01

The PRIDE (PyRoprocessing Integrated inactive DEmonstration) is an engineering-scale pyroprocessing test-bed facility that utilizes depleted uranium (DU) instead of spent fuel as a process material. As part of the ongoing effort to enhance pyroprocessing safeguardability, UNDA (Unified Non-Destructive Assay), a system integrating three different non-destructive assay techniques, namely, neutron, gamma-ray, and mass measurement, for nuclear material accountancy (NMA) was developed. In the present study, UNDA's NMA capability was evaluated by measurement of the weight, 238 U mass, and U enrichment of oxide-reduction-process feed material (i.e., porous pellets). In the 238 U mass determination, the total neutron counts for porous pellets of six different weights were measured. The U enrichment of the porous pellets, meanwhile, was determined according to the gamma spectrums acquired using UNDA's NaI-based enrichment measurement system. The results demonstrated that the UNDA system, after appropriate corrections, could be used in PRIDE NMA applications with reasonable uncertainty. It is expected that in the near future, the UNDA system will be tested with next-step materials such as the products of the oxide-reduction and electro-refining processes. Copyright © 2016 Elsevier Ltd. All rights reserved.

Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohd Fadzil, Syazwani Binti; Hrma, Pavel R.; Schweiger, Michael J.

Pyroprocessing is a reprocessing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl-KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the matrix at high loadings. Crystallization that occurs inmore » waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.« less

Project Report on Development of a Safeguards Approach for Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert Bean

The Idaho National Laboratory has undertaken an effort to develop a standard safeguards approach for international commercial pyroprocessing facilities. This report details progress for the fiscal year 2010 effort. A component by component diversion pathway analysis has been performed, and has led to insight on the mitigation needs and equipment development needed for a valid safeguards approach. The effort to develop an in-hot cell detection capability led to the digital cloud chamber, and more importantly, the significant potential scientific breakthrough of the inverse spectroscopy algorithm, including the ability to identify energy and spatial location of gamma ray emitting sources withmore » a single, non-complex, stationary radiation detector system. Curium measurements were performed on historical and current samples at the FCF to attempt to determine the utility of using gross neutron counting for accountancy measurements. A solid cost estimate of equipment installation at FCF has been developed to guide proposals and cost allocations to use FCF as a test bed for safeguards measurement demonstrations. A combined MATLAB and MCNPX model has been developed to perform detector placement calculations around the electrorefiner. Early harvesting has occurred wherein the project team has been requested to provide pyroprocessing technology and safeguards short courses.« less

Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

NASA Astrophysics Data System (ADS)

Mohd Fadzil, Syazwani; Hrma, Pavel; Schweiger, Michael J.; Riley, Brian J.

2015-10-01

Pyroprocessing is are processing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl-KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the glass matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.

Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Versey, Joshua R.

This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is basedmore » on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.« less

Characterization of Raw and Decopperized Anode Slimes from a Chilean Refinery

NASA Astrophysics Data System (ADS)

Melo Aguilera, Evelyn; Hernández Vera, María Cecilia; Viñals, Joan; Graber Seguel, Teófilo

2016-04-01

This work characterizes raw and decopperized slimes, with the objective of identifying the phases in these two sub-products. The main phases in copper anodes are metallic copper, including CuO, which are present in free form or associated with the presence of copper selenide or tellurides (Cu2(Se,Te)) and several Cu-Pb-Sb-As-Bi oxides. During electrorefining, the impurities in the anode release and are not deposited in the cathode, part of them dissolving and concentrated in the electrolyte, and others form a raw anode slime that contains Au, Ag, Cu, As, Se, Te and PGM, depending on the composition of the anode. There are several recovery processes, most of which involve acid leaching in the first step to dissolve copper, whose product is decopperized anode slime. SEM analysis revealed that the mineralogical species present in the raw anode slime under study were mainly eucarite (CuAgSe), naumannite (Ag2Se), antimony arsenate (SbAsO4), and lead sulfate (PbSO4). In the case of decopperized slime, the particles were mainly composed of SbAsO4 (crystalline appearance), non-stoichiometric silver selenide (Ag(2- x)Se), and chlorargyrite (AgCl).

US/UK Loan Account Project Status PMOD477

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevens, Patrice A.

2012-07-12

The viewgraphs describe the status of PMOD477 for LANL. The meeting will occur at DOE-HQ with NA-11 and Military Applications personnel in attendance. Serves to repatriate material with a balance to zero by December 2012. Phase 1 -- Establish formality of operations for War Reserve (WR): Complete surrogate taskings to A90 through a Materials Channel and perform US/UK lessons learned; Complete the US/UK agreed Quality Acceptance Plan, Materials Plan, Shipping procedure, and establish the formal UK/US point of contacts. Phase 2 -- Metal Manufacture (WR): Process material and store material as electrorefined metal (ER) rings, with initial assay and isotopicmore » analysis, prior to manufacturing. Material is cast into accepted configuration and appropriate acceptance document for each aliquot will be generated. Phase 3 -- Intermediate Material Manufacture, Packaging and Shipping (WR): Continue processing of the material in accepted configuration with appropriate acceptance documentation for each aliquot. Provide an initial tasking of the material owed to UK including appropriate quality acceptance documentation. Phase 4 -- Complete Tasking (WR). Phase 5 -- Residue Processing (Non-WR): Complete processing of residue material and waste into accepted configuration with appropriate acceptance document for disposal.« less

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

DOE PAGES

Riley, Brian J.; Pierce, David A.; Crum, Jarrod V.; ...

2017-09-22

The work presented here is based off the concept that the rare earth chloride (RECl 3) fission products within the used electrorefiner (ER) salt can be selectively removed as RECl 3 (not yet demonstrated) or precipitated out as a mixture of REOCl and REO x through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl 3s at 10 mass% into a 78%TeO 2-22%PbO glass while also showing that this same mixture of RECl 3s can be oxidized to REOCl at 300 °C and then to REO x by 1200 °C, evolvingmore » Cl 2(g). When the REO x mixture is heated at temperatures >1200 °C, the ratios of REO xs change. The mixture of REO x was then immobilized in a lanthanide borosilicate (LABS) glass at a high loading of 60 mass%. Both the 78%TeO 2-22%PbO glass and LABS glass systems show good chemical durability. In conclusion, the advantages and disadvantages of tellurite and LABS glasses are compared.« less

Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.; Benedict, Robert W.

The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses “throw-away” technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technologymore » developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology.« less

SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew C. Morrison; Kenneth J. Bateman; Michael F. Simpson

2010-11-01

ABSTRACT SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS Matthew C. Morrison, Kenneth J. Bateman, Michael F. Simpson Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 The ceramic waste process is the intended method for disposing of waste salt electrolyte, which contains fission products from the fuel-processing electrorefiners (ER) at the INL. When mixed and processed with other materials, the waste salt can be stored in a durable ceramic waste form (CWF). The development of the CWF has recently progressed from small-scale testing and characterization to full-scale implementation and experimentation using surrogate materialsmore » in lieu of the ER electrolyte. Two full-scale (378 kg and 383 kg) CWF test runs have been successfully completed with final densities of 2.2 g/cm3 and 2.1 g/cm3, respectively. The purpose of the first CWF was to establish material preparation parameters. The emphasis of the second pre-qualification test run was to evaluate a preliminary multi-section CWF container design. Other considerations were to finalize material preparation parameters, measure the material height as it consolidates in the furnace, and identify when cracking occurs during the CWF cooldown process.« less

Distillation and condensation of LiCl-KCl eutectic salts for a separation of pure salts from salt wastes from an electrorefining process

NASA Astrophysics Data System (ADS)

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

2009-12-01

Salt separation and recovery from the salt wastes generated from a pyrochemical process is necessary to minimize the high-level waste volumes and to stabilize a final waste form. In this study, the thermal behavior of the LiCl-KCl eutectic salts containing rare earth oxychlorides or oxides was investigated during a vacuum distillation and condensation process. LiCl was more easily vaporized than the other salts (KCl and LiCl-KCl eutectic salt). Vaporization characteristics of LiCl-KCl eutectic salts were similar to that of KCl. The temperature to obtain the vaporization flux (0.1 g min -1 cm -2) was decreased by much as 150 °C by a reduction of the ambient pressure from 5 Torr to 0.5 Torr. Condensation behavior of the salt vapors was different with the ambient pressure. Almost all of the salt vapors were condensed and were formed into salt lumps during a salt distillation at the ambient pressure of 0.5 Torr and they were collected in the condensed salt storage. However, fine salt particles were formed when the salt distillation was performed at 10 Torr and it is difficult for them to be recovered. Therefore, it is thought that a salt vacuum distillation and condensation should be performed to recover almost all of the vaporized salts at a pressure below 0.5 Torr.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Measurement of Irradiated Pyroprocessing Samples via Laser Induced Breakdown Spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phongikaroon, Supathorn

The primary objective of this research is to develop an applied technology and provide an assessment to remotely measure and analyze the real time or near real time concentrations of used nuclear fuel (UNF) dissolute in electrorefiners. Here, Laser-Induced Breakdown Spectroscopy (LIBS), in UNF pyroprocessing facilities will be investigated. LIBS is an elemental analysis method, which is based on the emission from plasma generated by focusing a laser beam into the medium. This technology has been reported to be applicable in the media of solids, liquids (includes molten metals), and gases for detecting elements of special nuclear materials. The advantagesmore » of applying the technology for pyroprocessing facilities are: (i) Rapid real-time elemental analysis|one measurement/laser pulse, or average spectra from multiple laser pulses for greater accuracy in < 2 minutes; (ii) Direct detection of elements and impurities in the system with low detection limits|element specific, ranging from 2-1000 ppm for most elements; and (iii) Near non-destructive elemental analysis method (about 1 g material). One important challenge to overcome is achieving high-resolution spectral analysis to quantitatively analyze all important fission products and actinides. Another important challenge is related to accessibility of molten salt, which is heated in a heavily insulated, remotely operated furnace in a high radiation environment with an argon atmosphere.« less

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, Eddie C.; Miller, William E.; Laidler, James J.

1997-01-01

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two.

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1997-07-22

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two. 12 figs.

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thornbury, M. L.; Juarez, C.; Krass, A. W.

Efforts are underway at the Y-12 National Security Complex (Y-12) to modernize the recovery, purification, and consolidation of un-irradiated, highly enriched uranium metal. Successful integration of advanced technology such as Electrorefining (ER) eliminates many of the intermediate chemistry systems and processes that are the current and historical basis of the nuclear fuel cycle at Y-12. The cost of operations, the inventory of hazardous chemicals, and the volume of waste are significantly reduced by ER. It also introduces unique material forms and compositions related to the chemistry of chloride salts for further consideration in safety analysis and engineering. The work hereinmore » briefly describes recent investigations of nuclear criticality for 235UO2Cl2 (uranyl chloride) and 6LiCl (lithium chloride) in aqueous solution. Of particular interest is the minimum critical mass of highly enriched uranium as a function of the molar ratio of 6Li to 235U. The work herein also briefly describes recent investigations of nuclear criticality for 235U metal reflected by salt mixtures of 6LiCl or 7LiCl (lithium chloride), KCl (potassium chloride), and 235UCl3 or 238UCl3 (uranium tri-chloride). Computational methods for analysis of nuclear criticality safety and published nuclear data are employed in the absence of directly relevant experimental criticality benchmarks.« less

Electrochemical studies and analysis of 1-10 wt% UCl3 concentrations in molten LiCl-KCl eutectic

NASA Astrophysics Data System (ADS)

Hoover, Robert O.; Shaltry, Michael R.; Martin, Sean; Sridharan, Kumar; Phongikaroon, Supathorn

2014-09-01

Three electrochemical methods - cyclic voltammetry (CV), chronopotentiometry (CP), and anodic stripping voltammetry (ASV) - were applied to solutions of up to 10 wt% UCl3 in the molten LiCl-KCl eutectic salt at 500 °C to determine electrochemical properties and behaviors and to help provide a scientific basis for the development of an in situ electrochemical probe for determining the concentration of uranium in a used nuclear fuel electrorefiner. Diffusion coefficients of UCl4 and UCl3 were calculated to be (6.72 ± 0.360) × 10-6 cm2/s and (1.04 ± 0.17) × 10-5 cm2/s, respectively. Apparent standard reduction potentials were determined to be (-0.381 ± 0.013) V and (-1.502 ± 0.076) V vs. 5 mol% Ag/AgCl or (-1.448 ± 0.013) V and (-2.568 ± 0.076) V vs. Cl2/Cl- for the U(IV)/U(III) and U(III)/U redox couples, respectively. In comparing this data with supercooled thermodynamic data to determine activity coefficients, the thermodynamic database used was important with resulting activity coefficients ranging from 2.34 × 10-3 to 1.08 × 10-2 for UCl4 and 4.94 × 10-5 to 4.50 × 10-4 for UCl3. Of anodic stripping voltammetry and cyclic voltammetry anodic or cathodic peaks, the CV cathodic peak height divided by square root of scan rate was shown to be the most reliable method of determining UCl3 concentration in the molten salt.

Absorption characteristics of anions (I-, Br-, and Te2-) into zeolite in molten LiCl-KCl eutectic salt

NASA Astrophysics Data System (ADS)

Uozumi, Koichi; Sugihara, Kei; Kinoshita, Kensuke; Koyama, Tadafumi; Tsukada, Takeshi; Terai, Takayuki; Suzuki, Akihiro

2014-04-01

The behaviors of anion fission product (FP) elements to be absorbed into zeolite in molten LiCl-KCl eutectic salt were studied using iodine, bromine, and tellurium. First, the type-A zeolite was selected as the most suitable type of zeolite among type-A, type-X, and type-Y zeolites through experiments in which zeolites were heated together with LiCl-KCl-KI salt. As the next step, experiments in which the type-A zeolite was immersed in molten LiCl-KCl salt containing various concentrations of iodine, bromine, or tellurium were performed. The degree of absorption of the anion FP elements was evaluated using the separation factor (SF) value versus chlorine. Although the SF values for iodine and tellurium were higher than 1.0, which meant that these elements were absorbed into the type-A zeolite more intensively than chlorine in the salt, the corresponding value for bromine was approximately 1.0. The effects of coexisting cation FPs were also examined using cesium, strontium, and neodymium, and it was revealed that the SF values for iodine were less than those in the case without cation addition. On the other hand, the SF values for tellurium were not affected by the coexistence of cesium and strontium. Finally, the feasibility of the present pyroprocess flowsheet was evaluated by calculating the inventory of each anion FP in an electrorefiner based on the obtained SF values instead of temporary values for the anion FPs absorption, which were set due to lack of experimental data.

Top Ten Reasons for DEOX as a Front End to Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; K.J. Bateman; S.D. Herrmann

A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less

Experimental Investigations into U/TRU Recovery using a Liquid Cadmium Cathode and Salt Containing High Rare Earth Concentrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shelly X. Li; Steven D. Herrmann; Michael F. Simpson

2009-09-01

Experimental Investigations into U/TRU Recovery using a Liquid Cadmium Cathode and Salt Containing High Rare Earth Concentrations Shelly X. Li, Steven D. Herrmann, and Michael F. Simpson Pyroprocessing Technology Department Idaho National Laboratory P.O. Box 1625, Idaho Falls, ID 83415 USA Abstract - A series of six bench-scale liquid cadmium cathode (LCC) tests was performed to obtain basic separation data with focus on the behavior of rare earth elements. The electrolyte used for the tests was a mixed salt from the Mk-IV and Mk-V electrorefiners, in which spent metal fuels from Experimental Breeder Reactor-II (EBR-II) had been processed. Rare earthmore » (RE) chlorides, such as NdCl3, CeCl3, LaCl3, PrCl3, SmCl3, and YCl3, were spiked into the salt prior to the first test to create an extreme case for investigating rare earth contamination of the actinides collected by a LCC. For the first two LCC tests, an alloy with the nominal composition of 41U-30Pu-5Am-3Np-20Zr-1RE was loaded into the anode baskets as the feed material. The anode feed material for Runs 3 to 6 was spent ternary fuel (U-19Pu-10Zr). The Pu/U ratio in the salt varied from 0.6 to 1.3. Chemical and radiochemical analytical results confirmed that U and transuranics can be collected into the LCC as a group under the given run conditions. The RE contamination level in the LCC product was up to 6.7 wt% of the total metal collected. The detailed data for partitioning of actinides and REs in the salt and Cd phases are reported in the paper.« less

Corrosion and Microstructure Correlation in Molten LiCl-KCl Medium

NASA Astrophysics Data System (ADS)

Ravi Shankar, A.; Mathiya, S.; Thyagarajan, K.; Kamachi Mudali, U.

2010-07-01

Pyrochemical reprocessing in molten chloride salt medium has been considered as one of the best options for the reprocessing of spent metallic fuels of future fast breeder reactors. The unit operations such as salt preparation, electrorefining, and cathode processing involve the presence of molten LiCl-KCl eutectic salt from 673 to 1373 K (400 to 1100 °C). The present work discusses the corrosion behavior of electroformed nickel (EF Ni) without and with nickel-tungsten (Ni-W) coating, 316L SS, and INCONEL 625 alloy in molten LiCl-KCl eutectic salt at 673 K, 773 K, and 873 K (400 °C, 500 °C, and 600 °C) in the presence of air. The weight percent loss of the exposed samples was determined by the weight loss method and surface morphology of the salt exposed, and product layers were examined by scanning electron microscopy (SEM). X-ray diffraction (XRD) and energy-dispersive X-ray (EDX) analysis were also carried out on the exposed and corrosion product layers to understand the phases present and the corrosion mechanism involved. The results of the present study indicated that INCONEL 625 alloy showed superior corrosion resistance compared to electroformed nickel (EF Ni), EF Ni with nickel-tungsten (Ni-W) coating (EF Ni-W), and 316L SS. The EF Ni with Ni-W coating exhibits better corrosion resistance than EF Ni without tungsten coating. Based on the surface morphology, XRD, and EDX analysis of corrosion product layers, the mechanism of corrosion of INCONEL 625 and 316L involves formation of chromium-rich compound at the surface and subsequent spallation. For the EF Ni, the porous thick NiO corrosion product allows the penetration of salt, thus accelerating the corrosion. Improved corrosion resistance of EF Ni-W was attributed to the W-rich NiO layer, while for INCONEL 625, the adherent and protective NiO layer improved the corrosion resistance. The article highlights the results of the present investigation.

Characterizing exposures to airborne metals and nanoparticle emissions in a refinery.

PubMed

Miller, Arthur; Drake, Pamela L; Hintz, Patrick; Habjan, Matt

2010-07-01

An air quality survey was conducted at a precious metals refinery in order to evaluate worker exposures to airborne metals and to provide detailed characterization of the aerosols. Two areas within the refinery were characterized: a furnace room and an electro-refining area. In line with standard survey practices, both personal and area air filter samples were collected on 37-mm filters and analyzed for metals by inductively coupled plasma-atomic emission spectroscopy. In addition to the standard sampling, measurements were conducted using other tools, designed to provide enhanced characterization of the workplace aerosols. The number concentration and number-weighted particle size distribution of airborne particles were measured with a fast mobility particle sizer (FMPS). Custom-designed software was used to correlate particle concentration data with spatial location data to generate contour maps of particle number concentrations in the work areas. Short-term samples were collected in areas of localized high concentrations and analyzed using transmission electron microscopy (TEM) and energy dispersive spectroscopy (EDS) to determine particle morphology and elemental chemistry. Analysis of filter samples indicated that all of the workers were exposed to levels of silver above the Occupational Safety and Health Administration permissible exposure limit of 0.01 mg m(-3) even though the localized ventilation was functioning. Measurements with the FMPS indicated that particle number concentrations near the furnace increased up to 1000-fold above the baseline during the pouring of molten metal. Spatial mapping revealed localized elevated particle concentrations near the furnaces and plumes of particles rising into the stairwells and traveling to the upper work areas. Results of TEM/EDS analyses confirmed the high number of nanoparticles measured by the FMPS and indicated the aerosols were rich in metals including silver, lead, antimony, selenium, and zinc. Results of the survey were used to deduce appropriate strategies for mitigation of worker exposure to airborne metals.

Development of High-Temperature Transport Technologies of Molten Salt Slurry in Pyrometallurgical Reprocessing

NASA Astrophysics Data System (ADS)

Hijikata, Takatoshi; Koyama, Tadafumi

Pyrometallurgical-reprocessing is one of the most promising technologies for advanced fuel cycle with favorable economic potential and intrinsic proliferation resistance. The development of transport technology for molten salt is a key issue in the industrialization of pyro-reprocessing. As for pure molten LiCl-KCl eutectic salt at approximately 773 K, we have already reported the successful results of transport using gravity and a centrifugal pump. However, molten salt in an electrorefiner mixes with insoluble fines when spent fuel is dissolved in porous anode basket. The insoluble consists of noble metal fission products, such as Pd, Ru, Mo, and Zr. There have been very few transport studies of a molten salt slurry (metal fines-molten salt mixture). Hence, transport experiments on a molten salt slurry were carried out to investigate the behavior of the slurry in a tube. The apparatus used in the transport experiments on the molten salt slurry consisted of a supply tank, a 10° inclined transport tube (10 mm inner diameter), a valve, a filter, and a recovery tank. Stainless steel (SS) fines with diameters from 53 to 415 μm were used. To disperse these fines homogenously, the molten salt and fines were stirred in the supply tank by an impeller at speeds from 1200 to 2100 rpm. The molten salt slurry containing 0.04 to 0.4 vol.% SS fines was transported from the supply tank to the recovery tank through the transportation tube. In the recovery tank, the fines were separated from the molten salt by the filter to measure the transport behavior of molten salt and SS fines. When the velocity of the slurry was 0.02 m/s, only 1% of the fines were transported to the recovery tank. On the other hand, most of the fines were transported when the velocity of the slurry was more than 0.8 m/s. Consequently, the molten salt slurry can be transported when the velocity is more than 0.8 m/s.

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

NASA Astrophysics Data System (ADS)

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.

Report on Concepts & Approaches for SSBD for eCHEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Chantell Lynne-Marie

The verification of special nuclear material (SNM) in spent fuel pyroprocessing is an important safeguards challenge. The detection of spontaneous fission (SF) neutrons from curium is an accepted, non-destructive technique that has been applied to verify special nuclear material (SNM) content in used fuel and other materials in the fuel cycle. The nuclear material accounting (NMA) technique at the Korea Atomic Energy Research Institute’s Reference Engineering-scale Pyroprocessing Facility (REPF) is based on the Cm balance technique. Several publications have demonstrated the safeguards benefit from using process monitoring (PM) on nuclear facilities as a complementary measure to NMA. More recently, thismore » concept was expanded and preliminarily demonstrated for pyroprocessing. The concept of Signature Based Safeguards (SBS) is part of this expansion, and is built around the interpretation of input from various sensors in a declared facility coupled with complementary NMA methods to increase confidence and lower standard error inventory differences (SEID). The SBS methodology was conceptually developed and relies on near real time analysis of process monitoring data to detect material diversion complemented by robust containment and surveillance (C/S) measures. This work demonstrates one example of how the SBS framework can be used in the electrorefiner. In this SBS application, a combination of cyclic voltammetry (CV) and neutron counting is applied to track and monitor Pu mass balance. The main purpose of this experiment is to determine if meaningful information can be gained from CV measurements with regard to the Mg/Gd ratio. This data will be coupled with ICP-MS to verify Gd concentrations and analyzed for statistical significance. It is expected the CV data will register a significant change under the off-normal operating conditions. Knowing how to identify and interpret those changes may help inform how to target more traditional neutron counting methods, which could support a more efficient safeguards system. The experimental results will be compared with theoretical calculations and the ERAD simulations.« less

Study of a double bubbler for material balance in liquids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hugues Lambert

The objective of this project was to determine the potential of a double bubbler to measure density and fluid level of the molten salt contained in an electrorefiner. Such in-situ real-time measurements can provide key information for material balances in the pyroprocessing of the nuclear spent fuel. This theoretical study showed this technique has a lot of promise. Four different experiments were designed and performed. The first three experiments studied the influence of a variety of factors such as depth difference between the two tubes, gas flow rate, the radius of the tubes and determining the best operating conditions. Themore » last experiment purpose was to determine the precision and accuracy of the apparatus during specific conditions. The elected operating conditions for the characterization of the system were a difference of depth of 25 cm and a flow rate of 55 ml/min in each tube. The measured densities were between 1,000 g/l and 1,400g/l and the level between 34cm and 40 cm. The depth difference between the tubes is critical, the larger, the better. The experiments showed that the flow rate should be the same in each tube. The concordances with theoretical predictions were very good. The density precision was very satisfying (spread<0.1%) and the accuracy was about 1%. For the level determination, the precision was also very satisfying (spread<0.1%), but the accuracy was about 3%. However, those two biases could be corrected with calibration curves. In addition to the aqueous systems studied in the present work, future work will focus on examining the behavior of the double bubbler instrumentation in molten salt systems. The two main challenges which were identified in this work are the effect of the temperature and the variation of the superficial tension.« less

Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hermann, S.D.; Gese, N.J.; Wurth, L.A.

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide.more » In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.« less

Automated design synthesis of robotic/human workcells for improved manufacturing system design in hazardous environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Joshua M.

Manufacturing tasks that are deemed too hazardous for workers require the use of automation, robotics, and/or other remote handling tools. The associated hazards may be radiological or nonradiological, and based on the characteristics of the environment and processing, a design may necessitate robotic labor, human labor, or both. There are also other factors such as cost, ergonomics, maintenance, and efficiency that also effect task allocation and other design choices. Handling the tradeoffs of these factors can be complex, and lack of experience can be an issue when trying to determine if and what feasible automation/robotics options exist. To address thismore » problem, we utilize common engineering design approaches adapted more for manufacturing system design in hazardous environments. We limit our scope to the conceptual and embodiment design stages, specifically a computational algorithm for concept generation and early design evaluation. In regard to concept generation, we first develop the functional model or function structure for the process, using the common 'verb-noun' format for describing function. A common language or functional basis for manufacturing was developed and utilized to formalize function descriptions and guide rules for function decomposition. Potential components for embodiment are also grouped in terms of this functional language and are stored in a database. The properties of each component are given as quantitative and qualitative criteria. Operators are also rated for task-relevant criteria which are used to address task compatibility. Through the gathering of process requirements/constraints, construction of the component database, and development of the manufacturing basis and rule set, design knowledge is stored and available for computer use. Thus, once the higher level process functions are defined, the computer can automate the synthesis of new design concepts through alternating steps of embodiment and function structure updates/decomposition. In the process, criteria guide function allocation of components/operators and help ensure compatibility and feasibility. Through multiple function assignment options and varied function structures, multiple design concepts are created. All of the generated designs are then evaluated based on a number of relevant evaluation criteria: cost, dose, ergonomics, hazards, efficiency, etc. These criteria are computed using physical properties/parameters of each system based on the qualities an engineer would use to make evaluations. Nuclear processes such as oxide conversion and electrorefining are utilized to aid algorithm development and provide test cases for the completed program. Through our approach, we capture design knowledge related to manufacturing and other operations in hazardous environments to enable a computational program to automatically generate and evaluate system design concepts.« less

Zone Freezing Study for Pyrochemical Process Waste Minimization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ammon Williams

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 hasmore » 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 temperature differences between the high and low furnace zones—200 and 300 ?C. During each experiment, the temperatures at selected locations around the crucible were measured and recorded to provide temperature profiles. Following each experiment, samples were collected and elemental analysis was done to determine the composition of iii the salt. Several models—non-mixed, well-mixed, Favier, and hybrid—were explored to describe the zone freezing process. For CsCl-LiCl-KCl system, experimental results indicate that through this process up to 90% of the used salt can be recycled, effectively reducing waste volume by a factor of ten. The optimal configuration was found to be a 5.0 mm/hr rate with a lid configuration and a ?T of 200°C. The larger 400 g mixtures had recycle percentages similar to the 50 g mixtures; however, the throughput per time was greater for the 400 g case. As a result, the 400 g case is recommended. For the CeCl3-LiCl-KCl system, the result implies that it is possible to use this process to separate the rare-earth and transuranics chlorides. Different models were applied to only CsCl ternary system. The best fit model was the hybrid model as a result of a solute transport transition from non- mixed to well-mixed throughout the growing process.« less

Fuel Cycle Research & Development Technical Monthly-March 2012

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Michael C.

2012-05-10

Several MPACT BCPs were executed in February, reflecting the shift in MPACT priorities directed late last year. Work continued on the FY2014 IPL, also bringing it in line with the new priorities. Preparations were made for the March MPACT Working Group meeting, in conjunction with Savannah River which is hosting the meeting. Steps were taken to initiate a new project with the World Institute for Nuclear Security, including discussions with WINS staff and preliminary work on the required procurement documentation. Several hardware issues were worked through. The newest detector array is working at LANL. A thorough analysis of previously collectedmore » Pu sample data using recently developed analysis code with improved spectral energy calibrations was completed. We now have a significantly better understanding of measurement uncertainties. Post-test analyses of the salt and sensor material for the first sensor test are almost complete. Sensor testing with different arrangements will continue and will be oriented based on post-test analysis of the first sensor test. Sensor materials for the next couple of tests are being fabricated. Materials with different annealing temperatures are being prepared for analysis. Fast Neutron Imaging to Quantify Nuclear Materials - The imager detectors repairs are complete and work with the imager is under way. The milestone requiring a report on LANSCE experiments was completed and submitted. Analysis of previous experiments and comparisons to simulations is near complete. Results are being compared with previous LANSCE-LSDS and RPI results. Additional data library (TENDL) is also being checked to see whether there are differences in the simulation results. The mid-year MIP Monitor project accomplishments and progress was presented at the MPACT meeting held in March at SRNL. Discussions around the meeting included inquiries into the feasibility of collecting process measurement data at H-Canyon, and it was explored further after the meeting. Kenneth Dayman, the graduate student from University of Texas, completed an initial draft of his master's thesis. His research will contribute to the multivariate classifier currently under development. Sarah Bender, the graduate student from Pennsylvania State University, presented her work on a poster and in a conference paper at the MARC IX meeting. A mass balance flowsheet for the fast reactor fuel was completed and a model simulation is scheduled to begin construction next month. The development of a mass balance flowsheet for light water reactor fuel will predict the behavior of the separation process using mathematical functions. The completed flowsheet will be utilized as the basis for constructing the model simulation for the electrochemical separations. Comments and review of the model from the MPACT Working Group meeting have been used to evaluate updates to the EChem model. A preliminary physical security layout has been developed in ATLAS. Thermal stability tests for high temperature microfluidic interconnections were completed on all compounds tested for bonding strength. An interconnection strategy was determined based on these results that we expect will allow for operation at 400C in the first generation of sampling systems. Design of the sampling system using the chosen interconnections was initiated, with handoff to an external foundry for fabrication based on ANL specified process conditions expected by the middle of the month. Monte Carlo simulations of the sampling system were conducted under conditions of realistic sampling size distributions, electrorefiner inhomogeneity distributions, and detector efficiencies. These simulations were used to establish a baseline limit of detection for system operation, assuming an on-line separation step is conducted before detection. Sensor for measuring density and depth of molten electrolyte - The procurement effort continued. 80% of the components ordered to assemble the double bubbler have arrived at the INL. Pratap Sadasivan, and his team have been working on the new metrics for proliferation and security. They have defined the basic structure and method, implementation strategy, needed data, and approach to application. Initial drafting of several sections of the milestone document was started. The MPACT Working Group meeting was hosted at SRS on March 13-15, 2012. Approximately 65 researchers from national labs, industry and universities attended the technical meetings at the Center for Hydrogen Research on March 13-14 with a working lunch each day. 37 persons participated in a site tour, including H-Canyon and the MOX Facility, on March 15. As part of the WG meeting, a presentation by SRNL was given on H-Canyon history, capabilities and opportunities for its use as an MPACT technology test bed. Used fuels storage security analysis, guidance and best practices - Coordination discussions continued for the MPACT used fuel security work packages.« less