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1

Process to remove actinides from soil using magnetic separation  

DOE Patents (OSTI)

A process of separating actinide-containing components from an admixture including forming a slurry including actinide-containing components within an admixture, said slurry including a dispersion-promoting surfactant, adjusting the pH of the slurry to within a desired range, and, passing said slurry through a pretreated matrix material, said matrix material adapted to generate high magnetic field gradients upon the application of a strong magnetic field exceeding about 0.1 Tesla whereupon a portion of said actinide-containing components are separated from said slurry and remain adhered upon said matrix material is provided.

Avens, Larry R. (Los Alamos, NM); Hill, Dallas D. (Los Alamos, NM); Prenger, F. Coyne (Los Alamos, NM); Stewart, Walter F. (Las Cruces, NM); Tolt, Thomas L. (Los Alamos, NM); Worl, Laura A. (Los Alamos, NM)

1996-01-01T23:59:59.000Z

2

512-S Facility, Actinide Removal Process Radiological Design Summary Report  

SciTech Connect

This report contains top-level requirements for the various areas of radiological protection for workers. Detailed quotations of the requirements for applicable regulatory documents can be found in the Radiological Design Summary Report Implementation Guide. For the purposes of demonstrating compliance with these requirements, per Engineering Standard 01064, ''shall consider / shall evaluate'' indicates that the designer must examine the requirement for the design and either incorporate or provide a technical justification as to why the requirement is not incorporated. This report describes how the Building 512-S, Actinide Removal Process meets the required radiological design criteria and requirements based on 10CFR835, DOE Order 420.1A, WSRC Manual 5Q and various other DOE guides and handbooks. The analyses supporting this Radiological Design Summary Report initially used a source term of 10.6 Ci/gallon of Cs-137 as the basis for bulk shielding calculations. As the project evolved, the source term was reduced to 1.1 Ci/gallon of Cs-137. This latter source term forms the basis for later dose rate evaluations.

Nathan, S.J.

2004-04-21T23:59:59.000Z

3

Actinide recovery process  

DOE Patents (OSTI)

Process for the removal of plutonium polymer and ionic actinides from aqueous solutions by absorption onto a solid extractant loaded on a solid inert support such as polystyrene-divinylbenzene. The absorbed actinides can then be recovered by incineration, by stripping with organic solvents, or by acid digestion. Preferred solid extractants are trioctylphosphine oxide and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and the like. 2 tabs.

Muscatello, A.C.; Navratil, J.D.; Saba, M.T.

1985-06-13T23:59:59.000Z

4

Actinide metal processing  

DOE Patents (OSTI)

A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

Sauer, N.N.; Watkin, J.G.

1992-03-24T23:59:59.000Z

5

Actinide metal processing  

DOE Patents (OSTI)

This invention is comprised of a process of converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is provided together with a low temperature process of preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

Sauer, N.N.; Watkin, J.G.

1991-04-05T23:59:59.000Z

6

A comparision of TRUEX and CMP solvent extraction processes for actinide removal from ICPP wastes  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP) is currently engaged in development efforts for the decontamination of high-level radioactive wastes generated from decades of nuclear fuel reprocessing. These wastes include several types of calcine, generated by high temperature solidification of reprocessing raffinates. In addition to calcine, there are smaller quantities of secondary wastes from decontamination and solvent wash activities which are typically referred to as sodium-bearing waste (SBW). Solvent extraction technologies based on octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO, the active extractant in the TRUEX process) and dihexyl-N,N-diethylcarbamoylmethylphosphonate (DHDECMP, the active extractant in the CMP process) are being evaluated for actinide partitioning from these waste streams. Calcines must first be dissolved in an appropriate acidic solution prior to treatment in solvent extraction based processes. The SBW is currently stored as an acidic solution and readily amenable to liquid extraction techniques. Development efforts to date have revolved around defining and refining baseline flowsheets with the TRUEX and CMP processes for each waste stream. Another objective of this work was to determine which of these technologies are best suited for the treatment of ICPP wastes. Laboratory batch contacts were performed to identify relevant chemistry and distribution coefficients. This information was then used to establish baseline flowsheet configuration with regard to chemistry. The laboratory data were used to model the behavior of the actinides and other constituents in the wastes in countercurrent, continuous processes based on centrifugal contactor technology. The laboratory data and modelling results form the basis for comparison of the two processes.

Herbst, R.S.; Brewer, K.N.; Garn, T.G.; Law, J.D. [and others

1996-04-01T23:59:59.000Z

7

Actinide removal from spent salts  

DOE Patents (OSTI)

A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

2002-01-01T23:59:59.000Z

8

Actinide removal from nitric acid waste streams  

SciTech Connect

Actinide separations research at the Rocky Flats Plant (RFP) has found ways to significantly improve plutonium secondary recovery and americium removal from nitric acid waste streams generated by plutonium purification operations. Capacity and breakthrough studies show anion exchange with Dowex 1x4 (50 to 100 mesh) to be superior for secondary recovery of plutonium. Extraction chromatography with TOPO(tri-n-octyl-phosphine oxide) on XAD-4 removes the final traces of plutonium, including hydrolytic polymer. Partial neutralization and solid supported liquid membrane transfer removes americium for sorption on discardable inorganic ion exchangers, potentially allowing for non-TRU waste disposal.

Muscatello, A.C.; Navratil, J.D.

1986-01-01T23:59:59.000Z

9

Bidentate organophosphorus solvent extraction process for actinide recovery and partition  

DOE Patents (OSTI)

A liquid-liquid extraction process for the recovery and partitioning of actinide values from acidic nuclear waste aqueous solutions, the actinide values including trivalent, tetravalent and hexavalent oxidation states is provided and includes the steps of contacting the aqueous solution with a bidentate organophosphorous extractant to extract essentially all of the actinide values into the organic phase. Thereafter the respective actinide fractions are selectively partitioned into separate aqueous solutions by contact with dilute nitric or nitric-hydrofluoric acid solutions. The hexavalent uranium is finally removed from the organic phase by contact with a dilute sodium carbonate solution.

Schulz, Wallace W. (Richland, WA)

1976-01-01T23:59:59.000Z

10

SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011  

Science Conference Proceedings (OSTI)

Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the inorganic species measured in the leachate do not exceed the MCL, SMCL or TW limits. (4) The inorganic waste species that exceeded the MCL by more than a factor of 10 were nitrate, nitrite and the sum of nitrate and nitrite. (5) Analyses met all quality assurance specifications of US EPA SW-846. (6) The organic species (benzene, toluene, 1-butanol, phenol) were either not detected or were less than reportable for the vault classification samples. (7) The gross alpha and radium isotopes could not be determined to the MCL because of the elevated background which raised the detection limits. (8) Most of the beta/gamma activity was from 137Cs and its daughter 137mBa. (9) The concentration of 137Cs and 90Sr were present in the leachate at concentrations 1/40th and 1/8th respectively than in the 2003 vault classification samples. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the SCHWMR R.61-79.261.24(b) RCRA metals requirements for a nonhazardous waste form. The TCLP leachate concentrations for nitrate, nitrite and the sum of nitrate and nitrite were greater than 10x the MCLs in SCDHEC Regulations R.61-107.19, Part I A, which confirms the Saltstone Disposal Facility classification as a Class 3 Landfill. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the R.61-79.268.48(a) non wastewater treatment standards.

Eibling, R.

2011-09-28T23:59:59.000Z

11

Actinide and lanthanide separation process (ALSEP)  

DOE Patents (OSTI)

The process of the invention is the separation of minor actinides from lanthanides in a fluid mixture comprising, fission products, lanthanides, minor actinides, rare earth elements, nitric acid and water by addition of an organic chelating aid to the fluid; extracting the fluid with a solvent comprising a first extractant, a second extractant and an organic diluent to form an organic extractant stream and an aqueous raffinate. Scrubbing the organic stream with a dicarboxylic acid and a chelating agent to form a scrubber discharge. The scrubber discharge is stripped with a simple buffering agent and a second chelating agent in the pH range of 2.5 to 6.1 to produce actinide and lanthanide streams and spent organic diluents. The first extractant is selected from bis(2-ethylhexyl)hydrogen phosphate (HDEHP) and mono(2-ethylhexyl)2-ethylhexyl phosphonate (HEH(EHP)) and the second extractant is selected from N,N,N,N-tetra-2-ethylhexyl diglycol amide (TEHDGA) and N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA).

Guelis, Artem V.

2013-01-15T23:59:59.000Z

12

Process for Making a Ceramic Composition for Immobilization of Actinides  

DOE Patents (OSTI)

Disclosed is a process for making a ceramic composition for the immobilization of actinides, particularly uranium and plutonium. The ceramic is a titanate material comprising pyrochlore, brannerite and rutile. The process comprises oxidizing the actinides, milling the oxides to a powder, blending them with ceramic precursors, cold pressing the blend and sintering the pressed material.

Ebbinghaus, Bartley B.; Van Konynenburg, Richard A.; Vance, Eric R.; Stewart, Martin W.; Walls, Philip A.; Brummond, William Allen; Armantrout, Guy A.; Curtis, Paul G.; Hobson, Beverly F.; Farmer, Joseph; Herman, Connie Cicero; Herman, David Thomas

1999-06-22T23:59:59.000Z

13

Process for making a ceramic composition for immobilization of actinides  

DOE Patents (OSTI)

Disclosed is a process for making a ceramic composition for the immobilization of actinides, particularly uranium and plutonium. The ceramic is a titanate material comprising pyrochlore, brannerite and rutile. The process comprises oxidizing the actinides, milling the oxides to a powder, blending them with ceramic precursors, cold pressing the blend and sintering the pressed material.

Ebbinghaus, Bartley B. (Livermore, CA); Van Konynenburg, Richard A. (Livermore, CA); Vance, Eric R. (Kirrawee, AU); Stewart, Martin W. (Barden Ridge, AU); Walls, Philip A. (Cronulla, AU); Brummond, William Allen (Livermore, CA); Armantrout, Guy A. (Livermore, CA); Herman, Connie Cicero (Pleasanton, CA); Hobson, Beverly F. (Livermore, CA); Herman, David Thomas (Pleasanton, CA); Curtis, Paul G. (Tracy, CA); Farmer, Joseph (Tracy, CA)

2001-01-01T23:59:59.000Z

14

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

Ackerman, John P. (Downers Grove, IL); Johnson, Terry R. (Wheaton, IL)

1994-01-01T23:59:59.000Z

15

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

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

1994-08-09T23:59:59.000Z

16

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

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

1992-01-01T23:59:59.000Z

17

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

SciTech Connect

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

Chiarizia, R.; Danesi, P.R.

1985-01-01T23:59:59.000Z

18

Silica Scaling Removal Process  

NLE Websites -- All DOE Office Websites (Extended Search)

Silica Scaling Removal Process Silica Scaling Removal Process Silica Scaling Removal Process Scientists at Los Alamos National Laboratory have developed a novel technology to remove both dissolved and colloidal silica using small gel particles. Available for thumbnail of Feynman Center (505) 665-9090 Email Silica Scaling Removal Process Applications: Cooling tower systems Water treatment systems Water evaporation systems Potential mining applications (produced water) Industry applications for which silica scaling must be prevented Benefits: Reduces scaling in cooling towers by up to 50% Increases the number of cycles of concentration substantially Reduces the amount of antiscaling chemical additives needed Decreases the amount of makeup water and subsequent discharged water (blowdown) Enables considerable cost savings derived from reductions in

19

Continuous sulfur removal process  

DOE Patents (OSTI)

A continuous process for the removal of hydrogen sulfide from a gas stream using a membrane comprising a metal oxide deposited on a porous support is disclosed. 4 figures.

Jalan, V.; Ryu, J.

1994-04-26T23:59:59.000Z

20

Citrate based ``TALSPEAK`` lanthanide-actinide separation process  

SciTech Connect

The potential hazard posed to future generations by long-lived radionuclides such as the transuranic elements (TRU) is perceived as a major problem associated with the use of nuclear power. TRU wastes have to remain isolated from the environment for ``geological`` periods of time. The costs of building, maintaining, and operating a ``geological TRU repository`` can be very high. Therefore, there are significant economical advantages in segregating the relatively low volume of TRU wastes from other nuclear wastes. The chemical behavior of lanthanides and actinides, 4f and 5f elements respectively, is rather similar. As a consequence, the separation of these two groups is difficult. The ``TALSPEAK`` process (Trivalent Actinide Lanthanide Separations by Phosphorus-reagent Extraction from Aqueous Complexes) is one of the few means available to separate the trivalent actinides from the lanthanides. The method is based on the preferential complexation of the trivalent actinides by an aminopolyacetic acid. Cold experiments showed that by using citric acid the deleterious effects produced by impurities such as zirconium are greatly reduced.

Del Cul, G.D.; Bond, W.D.; Toth, L.M.; Davis, G.D.; Dai, S.; Metcalf, D.H.

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Actinide partitioning from actual Idaho chemical processing plant acidic tank waste using centrifugal contactors  

Science Conference Proceedings (OSTI)

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of the actinides from acidic radioactive wastes stored at the ICPP. These efforts have culminated in a recent demonstration of the TRUEX process with actual tank waste. This demonstration was performed using 24 stages of 2-cm diameter centrifugal contactors installed in a shielded hot cell at the ICPP Remote Analytical Laboratory. An overall removal efficiency of 99.97% was obtained for the actinides. As a result, the activity of the actinides was reduced from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, which is well below the U.S. NRC Class A LLW requirement of 10 nCi/g for non-TRU waste. Iron was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Mercury was also extracted by the TRUEX solvent (76%) and stripped from the solvent in the 0.25 M Na{sub 2}CO{sub 3} wash section.

Law, J.D.; Brewer, K.N.; Todd, T.A.

1997-10-01T23:59:59.000Z

22

Flowsheet report for baseline actinide blanket processing for accelerator transmutation of waste  

Science Conference Proceedings (OSTI)

We provide a flowsheet analysis of the chemical processing of actinide and fission product materials form the actinide blanket of an accelerator-based transmutation concept. An initial liquid ion exchange step is employed to recover unburned plutonium and neptunium, so that it can be returned quickly to the transmitter. The remaining materials, consisting of fission products and trivalent actinides (americium, curium), is processed after a cooling period. A reverse Talspeak process is employed to separate these trivalent actinides from lanthanides and other fission products.

Walker, R.B.

1992-04-08T23:59:59.000Z

23

Key features of the Talspeak and similar trivalent actinide-lanthanide partitioning processes  

Science Conference Proceedings (OSTI)

As closing of the nuclear-fuel cycle via the suite of UREX processes under development in the U.S. progresses, the Trivalent Actinide-Lanthanide Separation by Phosphorus Extractants and Aqueous Komplexants (TALSPEAK) process has been selected as the baseline process for partition of trivalent actinides away from fission-product lanthanides. In this report, selected features of the chemistry of the TALSPEAK process and the limited parallel information on other TALSPEAK-like processes are discussed. (author)

Nash, Kenneth L. [Washington State University, Chemistry Department, P.O. Box 644630, Pullman, WA 99164-4630 (United States)

2008-07-01T23:59:59.000Z

24

Actinide Chemistry  

NLE Websites -- All DOE Office Websites (Extended Search)

sites. * Modeling actinide transport in subsurface environments. Enhancing the nuclear fuel cycle. Developing new separation processes to recycle waste. Better understanding...

25

Development and validation of process models for minor actinide separations processes using centrifugal contactors  

Science Conference Proceedings (OSTI)

As any future spent fuel treatment facility is likely to be based on intensified solvent extraction equipment it is important to understand the chemical and mass transfer kinetics of the processes involved. Two candidate minor actinide separations processes have been examined through a programme of modeling and experimental work to illustrate some of the issues to address in turning these technologies in to fully optimized processes suitable for industrialization. (authors)

Fox, O.D.; Carrott, M.J.; Gaubert, E.; Maher, C.J.; Mason, C.; Taylor, R.J.; Woodhead, D.A. [British Technology Centre, Nexia Solutions, Sellafield, Seascale, CA20 1PG (United Kingdom)

2007-07-01T23:59:59.000Z

26

Carbamoylmethylphosphoryl derivatives as actinide extractants: their significance in the processing and recovery of plutonium and other actinides  

SciTech Connect

Three classes of carbamoylmethylphosphoryl extractants were studied for their ability to extract selected tri-, tetra-, and hexavalent actinides from nitric acid. The three extractants are dihexyl-N,N-diethylcarbamoylmethylphosphonate (DHDECMP), hexyl hexyl-N,N-diethylcarbamoylmethylphosphinate (HHDECMP), and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide OphiD(IB)CMPO. The above three extractants were compared on the basis of nitric acid and extractant dependencies for Am(III), solubility of complexes on loading with Nd(III) and U(VI), and selectivity of actinide(III) over fission products. The influence of temperature on D/sub Am/ from LiNO/sub 3/ (10/sup -2/ M HNO/sub 3/) and from 3 M HNO/sub 3/ using dilute solutions of DHDECMP, HHDECMP OphiD(IB)CMPO in o-xylene showed that the increase in D/sub Am/ in the series phosphone-phophinate-phosphine oxide was due primarily to an increase in the enthalpy of extraction. This information was used to develop conceptual flowsheets for the extraction of all of the actinides (U, Np, Pu, Am, and Cm) from high-level waste from PUREX processing using 0.4 M OphiD(IB)CMPO in DEB and for the extraction of all of the actinides from dissolved spent LWR fuel using 0.8 M DHDECMP in DEB. In both flowsheets, no oxidation state of Pu is necessary since the III, IV, and VI state extract into the organic phase. 15 references, 6 figures, 7 tables.

Horwitz, E.P.; Diamond, H.; Kalina, D.G.

1983-01-01T23:59:59.000Z

27

Actinide solution processing at the Rocky Flats Environmental Technology Site  

SciTech Connect

The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1039, for radioactive solution removal and processing at Rocky Flats Environmental Technology Site, Golden, Colorado. The proposal for solution removal and processing is in response to independent safety assessments and an agreement with the State of Colorado to remove mixed residues at Rocky Flats and reduce the risk of future accidents. Monthly public meetings were held during the scoping and preparation of the EA. The scope of the EA included evaluations of alternative methods and locations of solution processing. A comment period from February 20, 1995 through March 21, 1995 was provided to the public and the State of Colorado to offer written comment on the EA. Comments were received from the State of Colorado and the U.S. Environmental Protection Agency. A response to the agency comments is included in the Final EA.

NONE

1995-04-01T23:59:59.000Z

28

Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste  

SciTech Connect

An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

1993-09-01T23:59:59.000Z

29

Toward understanding the thermodynamics of TALSPEAK process. Medium effects on actinide complexation  

SciTech Connect

The ingenious combination of lactate and diethylenetriamine-N,N,N’,N”,N”-pentaacetic acid (DTPA) as an aqueous actinide-complexing medium forms the basis of the successful separation of americium and curium from lanthanides known as the TALSPEAK process. While numerous reports in the prior literature have focused on the optimization of this solvent extraction system, considerably less attention has been devoted to the understanding of the basic thermodynamic features of the complex fluids responsible for the separation. The available thermochemical information of both lactate and DTPA protonation and metal complexation reactions are representative of the behavior of these ions under idealized conditions. Our previous studies of medium effects on lactate protonation suggest that significant departures from the speciation predicted based on reported thermodynamic values should be expected in the TALSPEAK aqueous environment. Thermodynamic parameters describing the separation chemistry of this process thus require further examination at conditions significantly removed from conventional ideal systems commonly employed in fundamental solution chemistry. Such thermodynamic characterization is the key to predictive modelling of TALSPEAK. Improved understanding will, in principle, allow process technologists to more efficiently respond to off-normal conditions during large scale process operation. In this report, the results of calorimetric and potentiometric investigations of the effects of aqueous electrolytes on the thermodynamic parameters for lactate protonation and lactate complexation of americium and neodymium will be presented. Studies on the lactate protonation equilibrium will clearly illustrate distinct thermodynamic variations between strong electrolyte aqueous systems and buffered lactate environment.

Peter R Zalupski; Leigh R Martin; Ken Nash; Yoshinobu Nakamura; Masahiko Yamamoto

2009-07-01T23:59:59.000Z

30

Dissolution of ORNL HLW sludge and partitioning of the actinides using the TRUEX process  

SciTech Connect

Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of {approximately} 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for {approximately} 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of {approximately} 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted.

Spencer, B.B.; Egan, B.Z.; Beahm, E.C.; Chase, C.W.; Dillow, T.A.

1997-12-01T23:59:59.000Z

31

Demonstration of the UNEX Process for the Simultaneous Separation of Cesium, Strontium, and the Actinides from Actual INEEL Tank Waste  

Science Conference Proceedings (OSTI)

A universal solvent extraction (UNEX) process for the simultaneous separation of cesium, strontium, and the actinides from actual radioactive acidic tank waste was demonstrated at the Idaho National Engineering and Environmental Laboratory. The waste solution used in the countercurrent flowsheet demonstration was obtained from tank WM-185. The UNEX process uses a tertiary solvent containing 0.08 M chlorinated cobalt dicarbollide, 0.5% polyethylene glycol-400 (PEG-400), and 0.02 M diphenyl-N,N-dibutylcarbamoyl phosphine oxide (Ph2Bu2CMPO) in a diluent consisting of phenyltrifluoromethyl sulfone (FS-13). The countercurrent flowsheet demonstration was performed in a shielded cell facility using 24 stages of 2-cm diameter centrifugal contactors. Removal efficiencies of 99.4%, 99.995%, and 99.96% were obtained for 137Cs, 90Sr, and total alpha, respectively. This is sufficient to reduce the activities of 137Cs, 90Sr, and actinides in the WM-185 waste to below NRC Class A LLW requirement s. Flooding and/or precipitate formation were not observed during testing. Significant amounts of the Zr (87%), Ba (>99%), Pb (98.8%), Fe (8%), Ca (10%), Mo (32%), and K (28%) were also removed from the feed with the universal solvent extraction flowsheet. 99Tc, Al, Hg, and Na were essentially inextractable (<1% extracted).

Law, J.D.; Herbst, R.S.; Todd, T.A. (INEEL); Romanovskiy, V.N.; Esimantovskiy, V.M.; Smirnov, I.V.; Babain, V.A.; Zaitsev, B.N. (V. G. Khlopin Radium Institute); Logunov, M.V. (MAYAK Production Association)

1999-10-01T23:59:59.000Z

32

Integrated AMP-PAN, TRUEX, and SREX Flowsheet Test to Remove Cesium, Surrogate Actinide Elements, and Strontium from INEEL Tank Waste Using Sorbent Columns and Centrifugal Contactors  

Science Conference Proceedings (OSTI)

Three unit operations for the removal of selected fission products, actinides, and RCRA metals (mercury and lead) have been successfully integrated and tested for extended run times with simulated INEEL acidic tank waste. The unit operations were ion exchange for Cs removal, followed by TRUEX solvent extraction for Eu (actinide surrogate), Hg, and Re (Tc surrogate) removal, and subsequent SREX solvent extraction for Sr and Pb removal. Approximately 45 L of simulated INTEC tank waste was first processed through three ion exchange columns in series for selective Cs removal. The columns were packed with a composite ammonium molybdophosphate-polyacrylonitrile (AMP-PAN) sorbent. The experimental breakthrough data were in excellent agreement with modeling predictions based on data obtained with much smaller columns. The third column (220 cm3) was used for polishing and Cs removal after breakthrough of the up-stream columns. The Cs removal was >99.83% in the ion exchange system without interference from other species. Most of the effluent from the ion exchange (IX) system was immediately processed through a TRUEX solvent extraction flowsheet to remove europium (americium surrogate), mercury and rhenium (technetium surrogate) from the simulated waste. The TRUEX flowsheet test was performed utilizing 23 stages of 3.3-cm centrifugal contactors. Greater than 99.999% of the Eu, 96.3% of the Hg, and 56% of the Re were extracted from the simulated feed and recovered in the strip and wash streams. Over the course of the test, there was no detectable build-up of any components in the TRUEX solvent. The raffinate from the TRUEX test was stored and subsequently processed several weeks later through a SREX solvent extraction flowsheet to remove strontium, lead, and Re (Tc surrogate) from the simulated waste. The SREX flowsheet test was performed using the same centrifugal contactors used in the TRUEX test after reconfiguration and the addition of three stages. Approximately 99.9% of the Sr, >99.89% of the Pb, and >96.4% of the Re were extracted from the aqueous feed to the SREX flowsheet and recovered in the strip and wash sections. Approximately 41 L of simulated tank waste (based on the volume processed through the TRUEX flowsheet) was processed through the integrated flowsheet and resulted in 175 L of liquid high activity waste (HAW) and 219.6 L of liquid low activity waste (LAW). The HAW fraction would be evaporated, dried and subsequently vitrified for final disposal. Based on current baseline assumptions, including a maximum phosphate loading of 2.5 wt. % in the HAW glass, the flowsheet tested would result in the production 0.195 kg of glass per L of tank waste processed. The LAW fraction would be solidified (via evaporation and denitration) and subsequently grouted. The current baseline assumptions for grouting the LAW stream indicate 0.37 kg of grout would be produced per L of tank waste treated. Under these assumptions, treating the current inventory of ~5E+6 L (5,000 m3) of tank waste would result in 375 m3 of HAW glass and 1,135 m3 of LAW Class A performance grout. The HAW glass volume could be significantly decreased by suitable TRUEX flowsheet modifications.

Herbst, Ronald Scott; Law, Jack Douglas; Todd, Terry Allen; Wood, D. J.; Garn, Troy Gerry; Wade, Earlen Lawrence

2000-02-01T23:59:59.000Z

33

Integrated AMP-PAN, TRUEX, and SREX Flowsheet Test to Remove Cesium, Surrogate Actinide Elements, and Strontium from INEEL Tank Waste Using Sorbent Columns and Centrifugal Contactors  

Science Conference Proceedings (OSTI)

Three unit operations for the removal of selected fission products, actinides, and RCRA metals (mercury and lead) have been successfully integrated and tested for extended run times with simulated INEEL acidic tank waste. The unit operations were ion exchange for Cs removal, followed by TRUEX solvent extraction for Eu (actinide surrogate), Hg, and Re (Tc surrogate) removal, and subsequent SREX solvent extraction for Sr and Pb removal. Approximately 45 L of simulated INTEC tank waste was first processed through three ion exchange columns in series for selective Cs removal. The columns were packed with a composite ammonium molybdophosphate-polyacrylonitrile (AMP-PAN) sorbent. The experimental breakthrough data were in excellent agreement with modeling predictions based on data obtained with much smaller columns. The third column (220 cm3) was used for polishing and Cs removal after breakthrough of the up-stream columns. The Cs removal was >99.83% in the ion exchange system without interference from other species. Most of the effluent from the ion exchange (IX) system was immediately processed through a TRUEX solvent extraction flowsheet to remove europium (americium surrogate), mercury and rhenium (technetium surrogate) from the simulated waste. The TRUEX flowsheet test was performed utilizing 23 stages of 3.3-cm centrifugal contactors. Greater than 99.999% of the Eu, 96.3% of the Hg, and 56% of the Re were extracted from the simulated feed and recovered in the strip and wash streams. Over the course of the test, there was no detectable build-up of any components in the TRUEX solvent. The raffinate from the TRUEX test was stored and subsequently processed several weeks later through a SREX solvent extraction flowsheet to remove strontium, lead, and Re (Tc surrogate) from the simulated waste. The SREX flowsheet test was performed using the same centrifugal contactors used in the TRUEX test after reconfiguration and the addition of three stages. Approximately 99.9% of the Sr, >99.89% of the Pb, and >96.4% of the Re were extracted from the aqueous feed to the SREX flowsheet and recovered in the strip and wash sections. Approximately 41 L of simulated tank waste (based on the volume processed through the TRUEX flowsheet) was processed through the integrated flowsheet and resulted in 175 L of liquid high activity waste (HAW) and 219.6 L of liquid low activity waste (LAW). The HAW fraction would be evaporated, dried and subsequently vitrified for final disposal. Based on current baseline assumptions, including a maximum phosphate loading of 2.5 wt. % in the HAW glass, the flowsheet tested would result in the production 0.195 kg of glass per L of tank waste processed. The LAW fraction would be solidified (via evaporation and denitration) and subsequently grouted. The current baseline assumptions for grouting the LAW stream indicate 0.37 kg of grout would be produced per L of tank waste treated. Under these assumptions, treating the current inventory of {approximately}5 E+6 L (5,000 m3) of tank waste would result in 375 m3 of HAW glass and 1,135 m3 of LAW Class A performance grout. The HAW glass volume could be significantly decreased by suitable TRUEX flowsheet modifications.

Herbst, R.S.; Law, J.D.; Todd, T.A.; Wood, D.J.; Garn, T.G.; Wade, E.L.

2000-01-31T23:59:59.000Z

34

DIAMIDE DERIVATIVES OF DIPICOLINIC ACID AS ACTINIDE AND LANTHANIDE EXTRACTANTS IN A VARIATION OF THE UNEX PROCESS  

SciTech Connect

The Universal Extraction (UNEX) process has been developed for simultaneous extraction of cesium, strontium, and actinides from acidic solutions. This process utilizes an extractant consisting of 0.08 M chlorinated cobalt dicarbollide (HCCD), 0.007-0.02 M polyethylene glycol (PEG-400), and 0.02 M diphenyl-N,N-di-n-butylcarbamoylmethylphosphine oxide (Ph2CMPO) in the diluent trifluoromethylphenyl sulfone (CF3C6H5SO2, designated FS-13) and provides simultaneous extraction of Cs, Sr, actinides, and lanthanides from HNO3 solutions. The UNEX process is of limited utility for processing acidic solutions containing large quantities of lanthanides and/or actinides, such as dissolved spent nuclear fuel solutions. These constraints are primarily attributed to the limited concentrations of CMPO (a maximum of ~0.02 M) in the organic phase and limited solubility of the CMPO-metal complexes. As a result, alternative actinide and lanthanide extractants are being investigated for use with HCCD as an improvement for waste processing and for applications where higher concentrations of the metals are present. Our preliminary results indicate that diamide derivatives of dipicolinic acid may function as efficient actinide and lanthanide extractants. The results to be presented indicate that, of the numerous diamides studied to date, the tetrabutyldiamide of dipicolinic acid, TBDPA, shows the most promise as an alternative actinide/lanthanide extractant in the UNEX process.

D. R. Peterman; R. S. Herbst; J. D. Law; R. D. Tillotson; T. G. Garn; T. A. Todd; V. N. Romanovskiy; V. A. Babain; M. Yu. Alyapyshev; I. V. Smirnov

2007-09-01T23:59:59.000Z

35

Improved sulfur removal processes evaluated for IGCC  

SciTech Connect

An inherent advantage of Integrated Coal Gasification Combined Cycle (IGCC) electric power generation is the ability to easily remove and recover sulfur. During the last several years, a number of new, improved sulfur removal and recovery processes have been commercialized. An assessment is given of alternative sulfur removal processes for IGCC based on the Texaco coal gasifier. The Selexol acid gas removal system, Claus sulfur recovery, and SCOT tail gas treating are currently used in Texaco-based IGCC. Other processes considered are: Purisol, Sulfinol-M, Selefning, 50% MDEA, Sulften, and LO-CAT. 2 tables.

1986-12-01T23:59:59.000Z

36

Catalyst regeneration process including metal contaminants removal  

DOE Patents (OSTI)

Spent catalysts removed from a catalytic hydrogenation process for hydrocarbon feedstocks, and containing undesired metals contaminants deposits, are regenerated. Following solvent washing to remove process oils, the catalyst is treated either with chemicals which form sulfate or oxysulfate compounds with the metals contaminants, or with acids which remove the metal contaminants, such as 5-50 W % sulfuric acid in aqueous solution and 0-10 W % ammonium ion solutions to substantially remove the metals deposits. The acid treating occurs within the temperature range of 60.degree.-250.degree. F. for 5-120 minutes at substantially atmospheric pressure. Carbon deposits are removed from the treated catalyst by carbon burnoff at 800.degree.-900.degree. F. temperature, using 1-6 V % oxygen in an inert gas mixture, after which the regenerated catalyst can be effectively reused in the catalytic process.

Ganguli, Partha S. (Lawrenceville, NJ)

1984-01-01T23:59:59.000Z

37

Selective Separation of Trivalent Actinides from Lanthanides by Aqueous Processing with Introduction of Soft Donor Atoms  

SciTech Connect

With increased application of MOX fuels and longer burnup times for conventional fuels, higher concentrations of the transplutonium actinides Am and Cm (and even heavier species like Bk and Cf) will be produced. The half-lives of the Am isotopes are significantly longer than those of the most important long-lived, high specific activity lanthanides or the most common Cm, Bk and Cf isotopes, thus the greatest concern as regards long-term radiotoxicity. With the removal and transmutation of Am isotopes, radiation levels of high level wastes are reduced to near uranium mineral levels within less than 1000 years as opposed to the time-fram if they remain in the wastes.

Kenneth L. Nash; Sue B. Clark; Gregg Lumetta

2009-09-23T23:59:59.000Z

38

Demonstration of the UNEX Process for the Simultaneous Separation of Cesium, Strontium, and the Actinides from Actual INEEL Sodium-Bearing Waste  

SciTech Connect

A universal solvent extraction (UNEX) process for the simultaneous separation of cesium, strontium, and the actinides from actual radioactive acidic tank waste was demonstrated at the Idaho National Engineering and Environmental Laboratory. The waste solution used in the countercurrent flowsheet demonstration was obtained from tank WM-185. The UNEX process uses a tertiary solvent containing 0.08 M chlorinated cobalt dicarbollide, 0.5% polyethylene glycol-400 (PEG-400), and 0.02 M diphenyl-N,N-dibutylcarbamoyl phosphine oxide (Ph2Bu2CMPO) in a diluent consisting of phenyltrifluoromethyl sulfone (FS-13). The countercurrent flowsheet demonstration was performed in a shielded cell facility using 24 stages of 2-cm diameter centrifugal contactors. Removal efficiencies of 99.4%, 99.995%, and 99.96% were obtained for 137Cs, 90Sr, and total alpha, respectively. This is sufficient to reduce the activities of 137Cs, 90Sr, and actinides in the WM-185 waste to below NRC Class A LLW requirements. Flooding and/or precipitate formation were not observed during testing. Significant amounts of the Zr (87%), Ba (>99%), Pb (98.8%), Fe (8%), Ca (10%), Mo (32%), and K (28%) were also removed from the feed with the universal solvent extraction flowsheet. 99Tc, Al, Hg, and Na were essentially inextractable (<1% extracted).

Law, Jack Douglas; Herbst, Ronald Scott; Todd, Terry Allen; Romanovskiy, V.; Smirnov, I.; Babain, V.; Zaitsev, B.; Esimantovskiy, V.

1999-11-01T23:59:59.000Z

39

Actinide Chemistry  

NLE Websites -- All DOE Office Websites (Extended Search)

Actinide Chemistry Actinide Chemistry Actinide Chemistry Research into alternative forms of energy, especially energy security, is one of the major national security imperatives of this century. Get Expertise David Gallimore Actinide Analytical Chemistry Email Rebecca Chamberlin Actinide Analytical Chemistry Email Josh Smith Chemistry Communications Email Along with the lanthanides, they are often called "the f-elements" because they have valence electrons in the f shell. Actinide chemistry serves a critical role in addressing global threats Project Description At Los Alamos, scientists are using actinide analytical chemistry to identify and quantify the chemical and isotopic composition of materials. Since the Manhattan Project, such work has supported the Laboratory's

40

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor  

SciTech Connect

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

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

1994-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Actinide separations by supported liquid membranes  

SciTech Connect

The work has demonstrated that actinide removal from synthetic waste solutions using both flat-sheet and hollow-fiber SLM's is a feasible chemical process at the laboratory scale level. The process is characterized by the typical features of SLM's processes: very small quantities of extractant required; the potential for operations with high feed/strip volume ratios, resulting in a corresponding concentration factor of the actinides; and simplicity of operation. Major obstacles to the implementation of the SLM technology to the decontamination of liquid nuclear wastes are the probable low resistance of polypropylene supports to high radiation fields, which may prevent the application to high-level nuclear wastes; the unknown lifetime of the SLM; and the high Na content of the separated actinide solution.

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

1984-01-01T23:59:59.000Z

42

Demonstration of the TRUEX process for partitioning of actinides from actual ICPP tank waste using centrifugal contactors in a shielded cell facility  

Science Conference Proceedings (OSTI)

TRUEX is being evaluated at Idaho Chemical Processing Plant (ICPP) for separating actinides from acidic radioactive waste stored at ICPP; efforts have culminated in a recent demonstration with actual tank waste. A continuous countercurrent flowsheet test was successfully completed at ICPP using waste from tank WM-183. This demonstration was performed using 24 states of 2-cm dia centrifugal contactors in the shielded hot cell at the ICPP Remote Analytical Laboratory. The flowsheet had 8 extraction stages, 5 scrub stages, 6 strip stages, 3 solvent wash stages, and 2 acid rinse stages. A centrifugal contactor stage in the scrub section was not working during testing, and the scrub feed (aqueous) solution followed the solvent into the strip section, eliminating the scrub section in the flowsheet. An overall removal efficiency of 99.97% was obtained for the actinides, reducing the activity from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, well below the NRC Class A LLW requirement of 10 nCi/g for non-TRU waste.The 0.04 M HEDPA strip section back-extracted 99.9998% of the actinide from the TRUEX solvent. Removal efficiencies of >99. 90, 99.96, 99.98, >98.89, 93.3, and 89% were obtained for {sup 241}Am, {sup 238}Pu, {sup 239}Pu, {sup 235}U, {sup 238}U, and {sup 99}Tc. Fe was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Hg was also extracted by the TRUEX solvent (73%) and stripped from the solvent in the 0.25 M Na2CO3 wash section. Only 1.4% of the Hg exited with the high activity waste strip product.

Law, J.D.; Brewer, K.N.; Herbst, R.S.; Todd, T.A.

1996-09-01T23:59:59.000Z

43

Design, synthesis, and evaluation of polyhydroxamate chelators for selective complexation of actinides  

SciTech Connect

Specific chelating polymers targeted for actinides have much relevance to problems involving remediation of nuclear waste. Goal is to develop polymer supported, ion specific extraction systems for removing actinides and other hazardous metal ions from wastewaters. This is part of an effort to develop chelators for removing actinide ions such as Pu from soils and waste streams. Selected ligands are being attached to polymeric backbones to create novel chelating polymers. These polymers and other water soluble and insoluble polymers have been synthesized and are being evaluated for ability to selectively remove target metal ions from process waste streams.

Gopalan, A.; Jacobs, H.; Koshti, N.; Stark, P.; Huber, V.; Dasaradhi, L.; Caswell, W. [New Mexico State Univ., Las Cruces, NM (United States); Smith, P.; Jarvinen, G. [Los Alamos National Lab., NM (United States)

1995-08-01T23:59:59.000Z

44

Process for removing metals from water  

DOE Patents (OSTI)

A process for removing metals from water including the steps of prefiltering solids from the water, adjusting the pH to between about 2 and 3, reducing the amount of dissolved oxygen in the water, increasing the pH to between about 6 and 8, adding water-soluble sulfide to precipitate insoluble sulfide- and hydroxide-forming metals, adding a containing floc, and postfiltering the resultant solution. The postfiltered solution may optionally be eluted through an ion exchange resin to remove residual metal ions. 2 tabs.

Napier, J.M.; Hancher, C.M.; Hackett, G.D.

1987-06-29T23:59:59.000Z

45

Combined Extraction of Cesium, Strontium, and Actinides from Alkaline Media: An Extension of the Caustic-Side Solvent Extraction (CSSX) Process Technology  

Science Conference Proceedings (OSTI)

The wastes present at DOE long-term storage sites are usually highly alkaline, and because of this, much of the actinides in these wastes are in the sludge phase. Enough actinide materials still remain in the supernatant liquid that they require separation followed by long-term storage in a geological repository. The removal of these metals from the liquid waste stream would permit their disposal as low-level waste and dramatically reduce the volume of high-level wastes.

Kenneth Raymond

2004-11-03T23:59:59.000Z

46

Transuranium Removal from Hanford AN-107 Simulants using Sodium Permanganate and Calcium  

SciTech Connect

Removal of strontium from the complexant-containing wastes (AN-102 and AN-107) had previously been acceptably accomplished by isotopic dilution. Actinide removal using ferric co-precipitation, however, was very problematic from both a processing and a decontamination standpoint. Therefore, a series of tests were performed to identify other potential actinide removal agents and to test these agents at various concentrations.

Wilmarth, W.

2000-08-30T23:59:59.000Z

47

Improved Processes to Remove Naphthenic Acids  

NLE Websites -- All DOE Office Websites (Extended Search)

Improved Processes to Remove Naphthenic Acids Improved Processes to Remove Naphthenic Acids Final Technical Report (From October 1, 2002 to September 30, 2005) Principle Authors Aihua Zhang, Qisheng Ma, Kangshi Wang, Yongchun Tang (co-PI), William A. Goddard (PI), Date Report was issued: December 9, 2005 DOE Award number: DE-FC26-02NT15383 Name and Address of Submitting Organization California Institute of Technology 1200 East California Blvd., Pasadena, CA91125 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any

48

Process for removing mercury from aqueous solutions  

DOE Patents (OSTI)

A process for removing mercury from water to a level not greater than two parts per billion wherein an anion exchange material that is insoluble in water is contacted first with a sulfide containing compound and second with a compound containing a bivalent metal ion forming an insoluble metal sulfide. To this treated exchange material is contacted water containing mercury. The water containing not more than two parts per billion of mercury is separated from the exchange material.

Googin, John M. (Oak Ridge, TN); Napier, John M. (Oak Ridge, TN); Makarewicz, Mark A. (Knoxville, TN); Meredith, Paul F. (Knoxville, TN)

1986-01-01T23:59:59.000Z

49

Process for removing mercury from aqueous solutions  

DOE Patents (OSTI)

A process for removing mercury from water to a level not greater than two parts per billion wherein an anion exchange material that is insoluble in water is contacted first with a sulfide containing compound and second with a compound containing a bivalent metal ion forming an insoluble metal sulfide. To this treated exchange material is contacted water containing mercury. The water containing not more than two parts per billion of mercury is separated from the exchange material.

Googin, J.M.; Napier, J.M.; Makarewicz, M.A.; Meredith, P.F.

1985-03-04T23:59:59.000Z

50

Process for removing sulfur from coal  

DOE Patents (OSTI)

A process is disclosed for the removal of divalent organic and inorganic sulfur compounds from coal and other carbonaceous material. A slurry of pulverized carbonaceous material is contacted with an electrophilic oxidant which selectively oxidizes the divalent organic and inorganic compounds to trivalent and tetravalent compounds. The carbonaceous material is then contacted with a molten caustic which dissolves the oxidized sulfur compounds away from the hydrocarbon matrix.

Aida, T.; Squires, T.G.; Venier, C.G.

1983-08-11T23:59:59.000Z

51

Demonstration of a TODGA/TBP process for recovery of trivalent actinides and lanthanides from a PUREX raffinate  

Science Conference Proceedings (OSTI)

The efficiency of the partitioning of trivalent actinides from a PUREX raffinate has been demonstrated with a TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. Based on the results coming from cold and hot batch extraction studies and with the aid of computer code calculations a continuous counter current process have been developed and two flowsheets were tested using miniature centrifugal contactors. The feed solutions was a synthetic PUREX raffinate, spiked with {sup 241}Am, {sup 244}Cm, {sup 252}Cf, {sup 152}Eu and {sup 134}Cs. More than 99.9 % of the trivalent actinides and lanthanides were extracted and back-extracted and very high decontamination factors to most fission products were obtained. Co-extraction of zirconium, molybdenum and palladium was prevented using oxalic acid and HEDTA. However 10% of ruthenium was extracted and only 3 % could be back extracted using diluted nitric acid. (authors)

Modolo, G.; Asp, H.; Vijgen, H. [Forschungszentrum Juelich GmbH, Institut fuer Energieforschung, 52425 Juelich (Germany); Malmbeck, R.; Magnusson, D. [European Commission, JRC, Institute for Transuranium Elements - ITU, 76125 Karlsruhe (Germany); Sorel, C. [Commissariat a l'Energie Atomique Valrho - CEA, DRCP/SCPS, BP17171, 30207 Bagnols-sur-Ceze (France)

2007-07-01T23:59:59.000Z

52

Process for removing polychlorinated biphenyls from soil  

DOE Patents (OSTI)

The present invention relates to a method of removing polychlorinated biphenyls from soil. The polychlorinated biphenyls are extracted from the soil by employing a liquid organic solvent dispersed in water in the ratio of about 1:3 to 3:1. The organic solvent includes such materials as short-chain hydrocarbons including kerosene or gasoline which are immiscible with water and are nonpolar. The organic solvent has a greater affinity for the PCB's than the soil so as to extract the PCB's from the soil upon contact. The organic solvent phase is separated from the suspended soil and water phase and distilled for permitting the recycle of the organic solvent phase and the concentration of the PCB's in the remaining organic phase. The present process can be satisfactorily practiced with soil containing 10 to 20% petroleum-based oils and organic fluids such as used in transformers and cutting fluids, coolants and the like which contain PCB's. The subject method provides for the removal of a sufficient concentration of PCB's from the soil to provide the soil with a level of PCB's within the guidelines of the Environmental Protection Agency.

Hancher, C.W.; Saunders, M.B.; Googin, J.M.

1984-11-16T23:59:59.000Z

53

Modified Bayer Process for Alumina Removal from Hanford Waste  

AREVA NC Inc. Modified Bayer Process for Alumina Removal from Hanford Waste January 24, 2007 Don Geniesse AREVA NC Inc.

54

IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS  

SciTech Connect

In the first year of this project, we have established our experimental and theoretical methodologies for studies of the catalytic decarboxylation process. We have developed both glass and stainless steel micro batch type reactors for the fast screening of various catalysts with reaction substrates of model carboxylic acid compounds and crude oil samples. We also developed novel product analysis methods such as GC analyses for organic acids and gaseous products; and TAN measurements for crude oil. Our research revealed the effectiveness of several solid catalysts such as NA-Cat-1 and NA-Cat-2 for the catalytic decarboxylation of model compounds; and NA-Cat-5{approx}NA-Cat-9 for the acid removal from crude oil. Our theoretical calculations propose a three-step concerted oxidative decarboxylation mechanism for the NA-Cat-1 catalyst.

Aihua Zhang; Qisheng Ma; William A. Goddard; Yongchun Tang

2004-04-28T23:59:59.000Z

55

PRODUCTION OF ACTINIDE METAL  

DOE Patents (OSTI)

A process of reducing actinide oxide to the metal with magnesium-zinc alloy in a flux of 5 mole% of magnesium fluoride and 95 mole% of magnesium chloride plus lithium, sodium, potassium, calcium, strontium, or barium chloride is presented. The flux contains at least 14 mole% of magnesium cation at 600-- 900 deg C in air. The formed magnesium-zinc-actinide alloy is separated from the magnesium-oxide-containing flux. (AEC)

Knighton, J.B.

1963-11-01T23:59:59.000Z

56

Improved Processes to Remove Naphthenic Acids  

Science Conference Proceedings (OSTI)

In the past three years, we followed the work plan as we suggested in the proposal and made every efforts to fulfill the project objectives. Based on our large amount of creative and productive work, including both of experimental and theoretic aspects, we received important technical breakthrough on naphthenic acid removal process and obtained deep insight on catalytic decarboxylation chemistry. In detail, we established an integrated methodology to serve for all of the experimental and theoretical work. Our experimental investigation results in discovery of four type effective catalysts to the reaction of decarboxylation of model carboxylic acid compounds. The adsorption experiment revealed the effectiveness of several solid materials to naphthenic acid adsorption and acidity reduction of crude oil, which can be either natural minerals or synthesized materials. The test with crude oil also received promising results, which can be potentially developed into a practical process for oil industry. The theoretical work predicted several possible catalytic decarboxylation mechanisms that would govern the decarboxylation pathways depending on the type of catalysts being used. The calculation for reaction activation energy was in good agreement with our experimental measurements.

Aihua Zhang; Qisheng Ma; Kangshi Wang; Yongchun Tang; William A. Goddard

2005-12-09T23:59:59.000Z

57

Actinides-1981  

Science Conference Proceedings (OSTI)

Abstracts of 134 papers which were presented at the Actinides-1981 conference are presented. Approximately half of these papers deal with electronic structure of the actinides. Others deal with solid state chemistry, nuclear physic, thermodynamic properties, solution chemistry, and applied chemistry.

Not Available

1981-09-01T23:59:59.000Z

58

Process for particulate removal from coal liquids  

DOE Patents (OSTI)

Suspended solid particulates are removed from liquefied coal products by first subjecting such products to hydroclone action for removal in the underflow of the larger size particulates, and then subjecting the overflow from said hydroclone action, comprising the residual finer particulates, to an electrostatic field in an electrofilter wherein such finer particulates are deposited in the bed of beads of dielectric material on said filter. The beads are periodically cleaned by backwashing to remove the accumulated solids.

Rappe, Gerald C. (Macungie, PA)

1983-01-01T23:59:59.000Z

59

Process for selected gas oxide removal by radiofrequency catalysts  

DOE Patents (OSTI)

This process to remove gas oxides from flue gas utilizes adsorption on a char bed subsequently followed by radiofrequency catalysis enhancing such removal through selected reactions. Common gas oxides include SO.sub.2 and NO.sub.x.

Cha, Chang Y. (3807 Reynolds St., Laramie, WY 82070)

1993-01-01T23:59:59.000Z

60

Process for removing technetium from iron and other metals  

DOE Patents (OSTI)

A process for removing technetium from iron and other metals comprises the steps of converting the molten, alloyed technetium to a sulfide dissolved in manganese sulfide, and removing the sulfide from the molten metal as a slag. 4 figs.

Leitnaker, J.M.; Trowbridge, L.D.

1999-03-23T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Process for removing technetium from iron and other metals  

DOE Patents (OSTI)

A process for removing technetium from iron and other metals comprises the steps of converting the molten, alloyed technetium to a sulfide dissolved in manganese sulfide, and removing the sulfide from the molten metal as a slag.

Leitnaker, James M. (Kingston, TN); Trowbridge, Lee D. (Oak Ridge, TN)

1999-01-01T23:59:59.000Z

62

Magnetic Process For Removing Heavy Metals From Water Employing Magnetites  

NLE Websites -- All DOE Office Websites (Extended Search)

Magnetic Process For Removing Heavy Metals From Water Employing Magnetic Process For Removing Heavy Metals From Water Employing Magnetites Magnetic Process For Removing Heavy Metals From Water Employing Magnetites A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. Available for thumbnail of Feynman Center (505) 665-9090 Email Magnetic Process For Removing Heavy Metals From Water Employing Magnetites A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. The magnetite is mixed with the water such that at least a portion of, and preferably the majority of, the heavy metal in the water is bound to the magnetite. Once this occurs the magnetite and

63

The Universal Solvent Exchange (UNEX) Process II: Flowsheet Development & Demonstration of the UNEX Process for the Separation of Cesium, Strontium, and Actinides from Actual Acidic Radioactive Waste  

Science Conference Proceedings (OSTI)

A novel solvent extraction process, the Universal Extraction (UNEX) process, has been developed for the simultaneous separation of cesium, strontium, and the actinides from acidic waste solutions. The UNEX process solvent consists of chlorinated cobalt dicarbollide for the extraction of 137Cs, polyethylene glycol for the extraction of 90Sr, and diphenyl-N,N-dibutylcarbamoyl phosphine oxide for the extraction of the actinides and lanthanides. A nonnitroaromatic polar diluent consisting of phenyltrifluoromethyl sulfone has been developed for this process. A UNEX flowsheet consisting of a single solvent extraction cycle has been developed as a part of a collaborative effort between the Khlopin Radium Institute (KRI) and the Idaho National Engineering and Environmental Laboratory (INEEL). This flowsheet has been demonstrated with actual acidic radioactive tank waste at the INEEL using 24 stages of 2-cm diameter centrifugal contactors installed in a shielded cell facility. The activities of 137Cs, 90Sr, and the actinides were reduced to levels at which a grout waste form would meet NRC Class A LLW requirements. The extraction of 99Tc and several nonradioactive metals by the UNEX solvent has also been evaluated.

Law, Jack Douglas; Herbst, Ronald Scott; Todd, Terry Allen; Romanovskiy, V. N.; Smirnov, I. V.; Esimantovskiy, V. M.; Zaitsev. B. N.; Babain, V. A.

2001-01-01T23:59:59.000Z

64

Process for removing carbon from uranium  

DOE Patents (OSTI)

Carbon contamination is removed from uranium and uranium alloys by heating in inert atmosphere to 700.degree.-1900.degree.C in effective contact with yttrium to cause carbon in the uranium to react with the yttrium. The yttrium is either in direct contact with the contaminated uranium or in indirect contact by means of an intermediate transport medium.

Powell, George L. (Oak Ridge, TN); Holcombe, Jr., Cressie E. (Knoxville, TN)

1976-01-01T23:59:59.000Z

65

IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS  

Science Conference Proceedings (OSTI)

In the second year of this project, we continued our effort to develop low temperature decarboxylation catalysts and investigate the behavior of these catalysts at different reaction conditions. We conducted a large number of dynamic measurements with crude oil and model compounds to obtain the information at different reaction stages, which was scheduled as the Task2 in our work plan. We developed a novel adsorption method to remove naphthenic acid from crude oil using naturally occurring materials such as clays. Our results show promise as an industrial application. The theoretical modeling proposed several possible reaction pathways and predicted the reactivity depending on the catalysts employed. From all of these studies, we obtained more comprehensive understanding about catalytic decarboxylation and oil upgrading based on the naphthenic acid removal concept.

Aihua Zhang; Qisheng Ma; Kangshi Wang, William A. Goddard, Yongchun Tang

2005-05-05T23:59:59.000Z

66

Defect processes involving oxygen-compensated sites in CaF/sub 2/ precipitates doped with lanthanides and actinides  

Science Conference Proceedings (OSTI)

Oxygen incorporation into calcium fluoride precipitates doped with lanthanides and actinides is investigated by use of the technique of site-selective spectroscopy. Fluorescence from erbium in specific fluoride- and oxygen-compensated sites is monitored as a function of the ignition temperature of the precipitate to study the conversion from fluoride to oxygen compensation. Another process, thermal annealing of a disordered precipitate to give a well-defined lattice, is also followed. Changes in both oxygen compensation formation and lattice annealing are found to occur upon the addition of other trivalent and monovalent ions. The results provide a better understanding of the solid-state chemistry involved in new methods of chemical analysis using rare-earth doped CaF/sub 2/ precipitates, and how certain interference effects can arise. Also included is a study of the temperature dependence of fluorescent sites in CaF/sub 2/:U/sup 6 +/.

Johnston, M.V.; Wright, J.C.

1981-10-15T23:59:59.000Z

67

Process for selected gas oxide removal by radiofrequency catalysts  

DOE Patents (OSTI)

This process to remove gas oxides from flue gas utilizes adsorption on a char bed subsequently followed by radiofrequency catalysis enhancing such removal through selected reactions. Common gas oxides include SO[sub 2] and NO[sub x]. 1 figure.

Cha, C.Y.

1993-09-21T23:59:59.000Z

68

Process for removing pyritic sulfur from bituminous coals  

DOE Patents (OSTI)

A process is provided for removing pyritic sulfur and lowering ash content of bituminous coals by grinding the feed coal, subjecting it to micro-agglomeration with a bridging liquid containing heavy oil, separating the microagglomerates and separating them to a water wash to remove suspended pyritic sulfur. In one embodiment the coal is subjected to a second micro-agglomeration step.

Pawlak, Wanda (Edmonton, CA); Janiak, Jerzy S. (Edmonton, CA); Turak, Ali A. (Edmonton, CA); Ignasiak, Boleslaw L. (Edmonton, CA)

1990-01-01T23:59:59.000Z

69

Thief Process Removal of Mercury from Flue Gas  

NLE Websites -- All DOE Office Websites (Extended Search)

Process for the Removal of Mercury from Flue Gas Process for the Removal of Mercury from Flue Gas Opportunity The Department of Energy's National Energy Technology Laboratory (NETL) is seeking licensing partners interested in implementing United States Patent Number 6,521,021 entitled "Thief Process for the Removal of Mercury from Flue Gas." Disclosed in this patent is a novel process in which partially combusted coal is removed from the combustion chamber of a power plant using a lance (called a "thief"). This partially combusted coal acts as a thermally activated adsorbent for mercury. When it is in- jected into the duct work of the power plant downstream from the exit port of the combustion chamber, mercury within the flue gas contacts and adsorbs onto the thermally activated sorbent. The sorbent-mercury

70

Process for removing cadmium from scrap metal  

DOE Patents (OSTI)

A process for the recovery of a metal, in particular, cadmium contained in scrap, in a stable form. The process comprises the steps of mixing the cadmium-containing scrap with an ammonium carbonate solution, preferably at least a stoichiometric amount of ammonium carbonate, and/or free ammonia, and an oxidizing agent to form a first mixture so that the cadmium will react with the ammonium carbonate to form a water-soluble ammine complex; evaporating the first mixture so that ammine complex dissociates from the first mixture leaving carbonate ions to react with the cadmium and form a second mixture that includes cadmium carbonate; optionally adding water to the second mixture to form a third mixture; adjusting the pH of the third mixture to the acid range whereby the cadmium carbonate will dissolve; and adding at least a stoichiometric amount of sulfide, preferably in the form of hydrogen sulfide or an aqueous ammonium sulfide solution, to the third mixture to precipitate cadmium sulfide. This mixture of sulfide is then preferably digested by heating to facilitate precipitation of large particles of cadmium sulfide. The scrap may be divided by shredding or breaking up to exposure additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal.

Kronberg, J.W.

1994-01-01T23:59:59.000Z

71

Process for removing cadmium from scrap metal  

DOE Patents (OSTI)

A process is described for the recovery of a metal, in particular, cadmium contained in scrap, in a stable form. The process comprises the steps of mixing the cadmium-containing scrap with an ammonium carbonate solution, preferably at least a stoichiometric amount of ammonium carbonate, and/or free ammonia, and an oxidizing agent to form a first mixture so that the cadmium will react with the ammonium carbonate to form a water-soluble ammine complex; evaporating the first mixture so that ammine complex dissociates from the first mixture leaving carbonate ions to react with the cadmium and form a second mixture that includes cadmium carbonate; optionally adding water to the second mixture to form a third mixture; adjusting the pH of the third mixture to the acid range whereby the cadmium carbonate will dissolve; and adding at least a stoichiometric amount of sulfide, preferably in the form of hydrogen sulfide or an aqueous ammonium sulfide solution, to the third mixture to precipitate cadmium sulfide. This mixture of sulfide is then preferably digested by heating to facilitate precipitation of large particles of cadmium sulfide. The scrap may be divided by shredding or breaking up to expose additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal. 2 figures.

Kronberg, J.W.

1995-04-11T23:59:59.000Z

72

Process for removing cadmium from scrap metal  

DOE Patents (OSTI)

A process for the recovery of a metal, in particular, cadmium contained in scrap, in a stable form. The process comprises the steps of mixing the cadmium-containing scrap with an ammonium carbonate solution, preferably at least a stoichiometric amount of ammonium carbonate, and/or free ammonia, and an oxidizing agent to form a first mixture so that the cadmium will react with the ammonium carbonate to form a water-soluble ammine complex; evaporating the first mixture so that ammine complex dissociates from the first mixture leaving carbonate ions to react with the cadmium and form a second mixture that includes cadmium carbonate; optionally adding water to the second mixture to form a third mixture; adjusting the pH of the third mixture to the acid range whereby the cadmium carbonate will dissolve; and adding at least a stoichiometric amount of sulfide, preferably in the form of hydrogen sulfide or an aqueous ammonium sulfide solution, to the third mixture to precipitate cadmium sulfide. This mixture of sulfide is then preferably digested by heating to facilitate precipitation of large particles of cadmium sulfide. The scrap may be divided by shredding or breaking up to expose additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal.

Kronberg, James W. (Aiken, SC)

1995-01-01T23:59:59.000Z

73

Decontamination of matrices containing actinide oxides  

DOE Patents (OSTI)

There is provided a method for removing actinides and actinide oxides, particularly fired actinides, from soil and other contaminated matrices, comprising: (a) contacting a contaminated material with a solution of at least one inhibited fluoride and an acid to form a mixture; (b) heating the mixture of contaminated material and solution to a temperature in the range from about 30 C to about 90 C while stirring; (c) separating the solution from any undissolved matrix material in the mixture; (d) washing the undissolved matrix material to remove any residual materials; and (e) drying and returning the treated matrix material to the environment.

Villarreal, Robert

1997-12-01T23:59:59.000Z

74

Process for removing heavy metal compounds from heavy crude oil  

DOE Patents (OSTI)

A process is provided for removing heavy metal compounds from heavy crude oil by mixing the heavy crude oil with tar sand; preheating the mixture to a temperature of about 650.degree. F.; heating said mixture to up to 800.degree. F.; and separating tar sand from the light oils formed during said heating. The heavy metals removed from the heavy oils can be recovered from the spent sand for other uses.

Cha, Chang Y. (Golden, CO); Boysen, John E. (Laramie, WY); Branthaver, Jan F. (Laramie, WY)

1991-01-01T23:59:59.000Z

75

Process for removing sulfur from sulfur-containing gases  

DOE Patents (OSTI)

The present disclosure relates to improved processes for treating hot sulfur-containing flue gas to remove sulfur therefrom. Processes in accorda The government may own certain rights in the present invention pursuant to EPA Cooperative Agreement CR 81-1531.

Rochelle, Gary T. (Austin, TX); Jozewicz, Wojciech (Chapel Hill, NC)

1989-01-01T23:59:59.000Z

76

Isotope Tracer Studies of Diffusion in Sillicates and of Geological Transport Processes Using Actinide Elements  

SciTech Connect

The objectives were directed toward understanding the transport of chemical species in nature, with particular emphasis on aqueous transport in solution, in colloids, and on particles. Major improvements in measuring ultra-low concentrations of rare elements were achieved. We focused on two areas of studies: (1) Field, laboratory, and theoretical studies of the transport and deposition of U, Th isotopes and their daughter products in natural systems; and (2) Study of calcium isotope fractionation effects in marine carbonates and in carbonates precipitated in the laboratory, under controlled temperature, pH, and rates of precipitation. A major study of isotopic fractionation of Ca during calcite growth from solution has been completed and published. It was found that the isotopic shifts widely reported in the literature and attributed to biological processes are in fact due to a small equilibrium fractionation factor that is suppressed by supersaturation of the solution. These effects were demonstrated in the laboratory and with consideration of the solution conditions in natural systems, where [Ca{sup 2+}] >> [CO{sub 3}{sup 2-}] + [HCO{sub 3}{sup -}]. The controlling rate is not the diffusion of Ca, as was earlier proposed, but rather the rate of supply of [CO{sub 3}{sup 2-}] ions to the interface. This now opens the issues of isotopic fractionation of many elements to a more physical-chemical approach. The isotopic composition of Ca {Delta}({sup 44}Ca/{sup 40}Ca) in calcite crystals has been determined relative to that in the parent solutions by TIMS using a double spike. Solutions were exposed to an atmosphere of NH{sub 3} and CO{sub 2}, provided by the decomposition of (NH4)2CO3. Alkalinity, pH, and concentrations of CO{sub 3}{sup 2-}, HCO{sub 3}{sup -}, and CO{sub 2} in solution were determined. The procedures permitted us to determine {Delta}({sup 44}Ca/{sup 40}Ca) over a range of pH conditions, with the associated ranges of alkalinity. Two solutions with greatly different Ca concentrations were used, but, in all cases, the condition [Ca] >> [CO{sub 3}{sup 2-}] was met. A wide range in {Delta}({sup 44}Ca/{sup 40}Ca) was found for the calcite crystals, extending from 0.04 {+-} 0.13 to -1.34 {+-} 0.15 {per_thousand}, generally anticorrelating with the amount of Ca removed from the solution. The results show that {Delta}({sup 44}Ca/{sup 40}Ca) is a linear function of the saturation state of the solution with respect to calcite ({Omega}). The two parameters are very well correlated over a wide range in {Omega} for each solution with a given [Ca]. Solutions, which were vigorously stirred, showed a much smaller range in {Delta}({sup 44}Ca/{sup 40}Ca) and gave values of -0.42 {+-} 0.14 {per_thousand}, with the largest effect at low {Omega}. It is concluded that the diffusive flow of CO{sub 3}{sup 2-} into the immediate neighborhood of the crystal-solution interface is the rate-controlling mechanism and that diffusive transport of Ca{sup 2+} is not a significant factor. The data are simply explained by the assumptions that: (a) the immediate interface of the crystal and the solution is at equilibrium with {Delta}({sup 44}Ca/{sup 40}Ca) {approx} -1.5 {+-} 0.25 {per_thousand}, and (b) diffusive inflow of CO{sub 3}{sup 2-} causes supersaturation, thus precipitating Ca from the regions, exterior to the narrow zone of equilibrium. We consider this model to be a plausible explanation of the available data reported in the literature. The well-resolved but small and regular isotope fractionation shifts in Ca are thus not related to the diffusion of very large hydrated Ca complexes, but rather due to the ready availability of Ca in the general neighborhood of the crystal solution interface. The largest isotopic shift which occurs is a small equilibrium effect which is then subdued by supersaturation precipitation for solutions where [Ca{sup 2+}] >> [CO{sub 3}{sup 2-}] + [HCO{sub 3}{sup -}]. It is shown that there is a clear temperature dependence of the net isotopic shifts, which is simply due to changes in {Omega}

Wasserburg, Gerald J

2008-07-31T23:59:59.000Z

77

Process for removal of sulfur oxides from waste gases  

Science Conference Proceedings (OSTI)

A process for removing sulfur oxides from waste gas is provided. The gas is contacted with a sorbent selected from sodium bicarbonate, trona and activated sodium carbonate and, utilizing an alkaline liquor containing borate ion so as to reduce flow rates and loss of alkalinity, the spent sorbent is regenerated with an alkaline earth metal oxide or hydroxide.

Lowell, P.S.; Phillips, J.L.

1983-05-24T23:59:59.000Z

78

Process for Removing Radioactive Wastes from Liquid Streams  

SciTech Connect

The process is under development at Mound Laboratory to remove radioactive waste (principally plutonium-238) from process water prior to discharge of the water to the Miami river. The contaminated water, as normally received, is at a pH between 6 and 90. Under these conditions, plutonium in all its oxidation states is hydrolyzed; however, the level of the radioactive solids varies from about 50ppm down to about 50 ppb and the plutonium remains in a colloidal or subcolloidal condition. The permissible concentration for discharge to the river is about 50 parts per trillion. Pilot plant test show that 95-99% of the radioactive material is removed by adsorption on diatomaceous earth. The remainder is removed by passage through a bed of either dibasic or tribasic calcium phosphate. Ground phosphate rock is equally effective in removing the radioactive material if the flow rate is controlled to permit sufficient contact time. Parameters for optimizing the process are now under study. Future plans include application of the process to wastes from reactor fuels reprocessing.

Kirby, H. W.; Blane, D. E.; Smolin, R. I.

1972-10-01T23:59:59.000Z

79

Actinide halide complexes  

DOE Patents (OSTI)

A compound of the formula MX{sub n}L{sub m} wherein M = Th, Pu, Np,or Am thorium, X = a halide atom, n = 3 or 4, L is a coordinating ligand selected from the group consisting of aprotic Lewis bases having an oxygen-, nitrogen-, sulfur-, or phosphorus-donor, and m is 3 or 4 for monodentate ligands or is 2 for bidentate ligands, where n + m = 7 or 8 for monodentate ligands or 5 or 6 for bidentate ligands, a compound of the formula MX{sub n} wherein M, X, and n are as previously defined, and a process of preparing such actinide metal compounds including admixing the actinide metal in an aprotic Lewis base as a coordinating solvent in the presence of a halogen-containing oxidant, are provided.

Avens, L.R.; Zwick, B.D.; Sattelberger, A.P.; Clark, D.L.; Watkin, J.G.

1991-02-07T23:59:59.000Z

80

More Economical Sulfur Removal for Fuel Processing Plants  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

enabled TDA to develop and commercialize its direct oxidation process-a simple, catalyst-based system for removing sulfur from natural gas and petroleum-that was convenient and economical enough for smaller fuel processing plants to use. TDA Research, Inc. (TDA) of Wheat Ridge, CO, formed in 1987, is a privately-held R&D company that brings products to market either by forming internal business

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Process for removing an organic compound from water  

DOE Patents (OSTI)

A process for removing organic compounds from water is disclosed. The process involves gas stripping followed by membrane separation treatment of the stripping gas. The stripping step can be carried out using one or multiple gas strippers and using air or any other gas as stripping gas. The membrane separation step can be carried out using a single-stage membrane unit or a multistage unit. Apparatus for carrying out the process is also disclosed. The process is particularly suited for treatment of contaminated groundwater or industrial wastewater.

Baker, Richard W. (Palo Alto, CA); Kaschemekat, Jurgen (Palo Alto, CA); Wijmans, Johannes G. (Menlo Park, CA); Kamaruddin, Henky D. (San Francisco, CA)

1993-12-28T23:59:59.000Z

82

33rd Actinide Separations Conference  

SciTech Connect

Welcome to the 33rd Actinide Separations Conference hosted this year by the Lawrence Livermore National Laboratory. This annual conference is centered on the idea of networking and communication with scientists from throughout the United States, Britain, France and Japan who have expertise in nuclear material processing. This conference forum provides an excellent opportunity for bringing together experts in the fields of chemistry, nuclear and chemical engineering, and actinide processing to present and discuss experiences, research results, testing and application of actinide separation processes. The exchange of information that will take place between you, and other subject matter experts from around the nation and across the international boundaries, is a critical tool to assist in solving both national and international problems associated with the processing of nuclear materials used for both defense and energy purposes, as well as for the safe disposition of excess nuclear material. Granlibakken is a dedicated conference facility and training campus that is set up to provide the venue that supports communication between scientists and engineers attending the 33rd Actinide Separations Conference. We believe that you will find that Granlibakken and the Lake Tahoe views provide an atmosphere that is stimulating for fruitful discussions between participants from both government and private industry. We thank the Lawrence Livermore National Laboratory and the United States Department of Energy for their support of this conference. We especially thank you, the participants and subject matter experts, for your involvement in the 33rd Actinide Separations Conference.

McDonald, L M; Wilk, P A

2009-05-04T23:59:59.000Z

83

Metal chelate process to remove pollutants from fluids  

DOE Patents (OSTI)

The present invention relates to improved methods using an organic iron chelate to remove pollutants from fluids, such as flue gas. Specifically, the present invention relates to a process to remove NO[sub x] and optionally SO[sub 2] from a fluid using a metal ion (Fe[sup 2+]) chelate wherein the ligand is a dimercapto compound wherein the --SH groups are attached to adjacent carbon atoms (HS--C--C--SH) or (SH--C--CCSH) and contain a polar functional group so that the ligand of DMC chelate is water soluble. Alternatively, the DMC is covalently attached to a water insoluble substrate such as a polymer or resin, e.g., polystyrene. The chelate is regenerated using electroreduction or a chemical additive. The dimercapto compound bonded to a water insoluble substrate is also useful to lower the concentration or remove hazardous metal ions from an aqueous solution. 26 figures.

Chang, S.G.T.

1994-12-06T23:59:59.000Z

84

Purex Process Improvements for Pu and NP Control in Total Actinide Recycle Flowsheets  

Science Conference Proceedings (OSTI)

Significant improvements are required in the Purex process to optimise it for Advanced Fuel Cycles. Two key challenges we have identified are, firstly, developing more efficient methods for U/Pu separations especially at elevated Pu concentrations and, secondly, improving recovery, control and routing of Np in a modified Purex process. A series of Purex-like flowsheets for improved Pu separations based on hydroxamic acids and are reported. Purex-like flowsheets have been tested on a glovebox-housed 30-stage miniature centrifugal contactor train. A series of trials have been performed to demonstrate the processing of feeds with varying Pu contents ranging from 7 - 40% by weight. These flowsheets have demonstrated hydroxamic acids are excellent reagents for complexant stripping of Pu being able to achieve high decontamination factors (DF) on both the U and Pu product streams and co - recover Np with Pu. The advantages of a complexant-based approach are shown to be especially relevant when AFC scenarios are considered, where the Pu content of the fuel is expected to b e significantly higher. Recent results towards modifying the Purex process to improve recovery and control of Np in short residence time contactors are reported. Work on the development of chemical and process models to describe the complicated behaviour of Np under primary separation conditions (i.e. the HA extraction contactor) is described. To test the performance of the model a series of experiments were performed including testing of flowsheets on a fume-hood housed miniature centrifugal contactor train. The flowsheet was designed to emulate the conditions of a primar y separations contactor with the Np split between the U-solvent product and aqueous raffinate. In terms of Np routing the process model showed good agreement with flowsheet trial however much further work is required to fully understand this complex system. (authors)

Birkett, J.E.; Carrott, M.J.; Crooks, G.; Fox, O.D.; Maher, C.J.; Taylor, R.J.; Woodhead, D.A. [Nexia Solutions Ltd., BNFL, Sellafield, Seascale, CA20 1PG (United Kingdom)

2006-07-01T23:59:59.000Z

85

The Thief Process for Mercury Removal from Flue Gas  

E-Print Network (OSTI)

The Thief Process is a cost-effective variation to activated carbon injection (ACI) for removal of mercury from flue gas. In this scheme, partially combusted coal from the furnace of a pulverized coal power generation plant is extracted by a lance and then re-injected into the ductwork downstream of the air preheater. Recent results on a 500-lb/hr pilot-scale combustion facility show similar removals of mercury for both the Thief Process and ACI. The tests conducted to date at laboratory, bench, and pilot-scales demonstrate that the Thief sorbents exhibit capacities for mercury from flue gas streams that are comparable to those exhibited by commercially available activated carbons. Independent verification of the sorbent activity at a pilot-plant that uses a slipstream from a Wisconsin utility has been accomplished. A patent for the process was issued in February 2003 [1]. The Thief sorbents are cheaper than commerciallyavailable activated carbons; exhibit excellent capacities for mercury; and the overall process holds great potential for reducing the cost of mercury removal from flue gas [1-4].

Evan J. Granite; Mark C. Freeman; Richard A. Hargis; William J. O’dowd; Henry W. Pennline

2004-01-01T23:59:59.000Z

86

Process for removing technetium from iron and other metals  

DOE Patents (OSTI)

Technetium is a radioactive product of the nuclear fission process. During reprocessing of spent or partially spent fuel from nuclear reactors, the technetium can be released and contaminate other, otherwise good, metals. A specific example is equipment in gaseous diffusion uranium enrichment cascades which have been used to process fuel which was returned from reactors, so-called reactor returns. These returns contained volatile technetium compounds which contaminated the metals in the equipment. Present regulations require that technetium be removed before the metal can be re-used at non-radioactive sites. Removing the technetium from contaminated metals has two desirable results. First, the large amount of nonradioactive metal produced by the process herein described can be recycled at a much lower cost than virgin metal can be produced. Second, large amounts of radioactively contaminated metal can be reduced to relatively small amounts of radioactive slag and large amounts of essentially uncontaminated metal. A new and improved process for removing technetium from iron and other metals is described in which between 1/10 atom % and 5 atom % of manganese is added to the contaminated metal in order to replace the technetium.

Leitnaker, James M.; Trowbridge, Lee D.

1997-12-01T23:59:59.000Z

87

Process for removing sulfate anions from waste water  

DOE Patents (OSTI)

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

Nilsen, David N. (Lebanon, OR); Galvan, Gloria J. (Albany, OR); Hundley, Gary L. (Corvallis, OR); Wright, John B. (Albany, OR)

1997-01-01T23:59:59.000Z

88

Removal of mercury from coal via a microbial pretreatment process  

Science Conference Proceedings (OSTI)

A process for the removal of mercury from coal prior to combustion is disclosed. The process is based on use of microorganisms to oxidize iron, sulfur and other species binding mercury within the coal, followed by volatilization of mercury by the microorganisms. The microorganisms are from a class of iron and/or sulfur oxidizing bacteria. The process involves contacting coal with the bacteria in a batch or continuous manner. The mercury is first solubilized from the coal, followed by microbial reduction to elemental mercury, which is stripped off by sparging gas and captured by a mercury recovery unit, giving mercury-free coal. The mercury can be recovered in pure form from the sorbents via additional processing.

Borole, Abhijeet P. (Knoxville, TN); Hamilton, Choo Y. (Knoxville, TN)

2011-08-16T23:59:59.000Z

89

Extraction processes and solvents for recovery of cesium, strontium, rare earth elements, technetium and actinides from liquid radioactive waste  

DOE Patents (OSTI)

Cesium and strontium are extracted from aqueous acidic radioactive waste containing rare earth elements, technetium and actinides, by contacting the waste with a composition of a complex organoboron compound and polyethylene glycol in an organofluorine diluent mixture. In a preferred embodiment the complex organoboron compound is chlorinated cobalt dicarbollide, the polyethylene glycol has the formula RC.sub.6 H.sub.4 (OCH.sub.2 CH.sub.2).sub.n OH, and the organofluorine diluent is a mixture of bis-tetrafluoropropyl ether of diethylene glycol with at least one of bis-tetrafluoropropyl ether of ethylene glycol and bis-tetrafluoropropyl formal. The rare earths, technetium and the actinides (especially uranium, plutonium and americium), are extracted from the aqueous phase using a phosphine oxide in a hydrocarbon diluent, and reextracted from the resulting organic phase into an aqueous phase by using a suitable strip reagent.

Zaitsev, Boris N. (St. Petersburg, RU); Esimantovskiy, Vyacheslav M. (St. Petersburg, RU); Lazarev, Leonard N. (St. Petersburg, RU); Dzekun, Evgeniy G. (Ozersk, RU); Romanovskiy, Valeriy N. (St. Petersburg, RU); Todd, Terry A. (Aberdeen, ID); Brewer, Ken N. (Arco, ID); Herbst, Ronald S. (Idaho Falls, ID); Law, Jack D. (Pocatello, ID)

2001-01-01T23:59:59.000Z

90

Actinide separations conference  

Science Conference Proceedings (OSTI)

This report contains the abstracts for 55 presentations given at the fourteenth annual Actinide Separations Conference. (JDL)

Not Available

1990-01-01T23:59:59.000Z

91

RAPID SEPARATION METHOD FOR ACTINIDES IN EMERGENCY SOIL SAMPLES  

Science Conference Proceedings (OSTI)

A new rapid method for the determination of actinides in soil and sediment samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used for samples up to 2 grams in emergency response situations. The actinides in soil method utilizes a rapid sodium hydroxide fusion method, a lanthanum fluoride soil matrix removal step, and a streamlined column separation process with stacked TEVA, TRU and DGA Resin cartridges. Lanthanum was separated rapidly and effectively from Am and Cm on DGA Resin. Vacuum box technology and rapid flow rates are used to reduce analytical time. Alpha sources are prepared using cerium fluoride microprecipitation for counting by alpha spectrometry. The method showed high chemical recoveries and effective removal of interferences. This new procedure was applied to emergency soil samples received in the NRIP Emergency Response exercise administered by the National Institute for Standards and Technology (NIST) in April, 2009. The actinides in soil results were reported within 4-5 hours with excellent quality.

Maxwell, S.; Culligan, B.; Noyes, G.

2009-11-09T23:59:59.000Z

92

Process for removal of sulfur compounds from fuel gases  

DOE Patents (OSTI)

Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

Moore, Raymond H. (Richland, WA); Stegen, Gary E. (Richland, WA)

1978-01-01T23:59:59.000Z

93

Extraction process for removing metallic impurities from alkalide metals  

DOE Patents (OSTI)

A development is described for removing metallic impurities from alkali metals by employing an extraction process wherein the metallic impurities are extracted from a molten alkali metal into molten lithium metal due to the immiscibility of the alkali metals in lithium and the miscibility of the metallic contaminants or impurities in the lithium. The purified alkali metal may be readily separated from the contaminant-containing lithium metal by simple decanting due to the differences in densities and melting temperatures of the alkali metals as compared to lithium.

Royer, Lamar T. (Knoxville, TN)

1988-01-01T23:59:59.000Z

94

Extraction process for removing metallic impurities from alkalide metals  

DOE Patents (OSTI)

A development is described for removing metallic impurities from alkali metals by employing an extraction process wherein the metallic impurities are extracted from a molten alkali metal into molten lithium metal due to the immiscibility of the alkali metals in lithium and the miscibility of the metallic contaminants or impurities in the lithium. The purified alkali metal may be readily separated from the contaminant-containing lithium metal by simple decanting due to the differences in densities and melting temperatures of the alkali metals as compared to lithium.

Royer, L.T.

1987-03-20T23:59:59.000Z

95

Process for removal of hazardous air pollutants from coal  

SciTech Connect

An improved process for removing mercury and other trace elements from coal containing pyrite by forming a slurry of finely divided coal in a liquid solvent capable of forming ions or radicals having a tendency to react with constituents of pyrite or to attack the bond between pyrite and coal and/or to react with mercury to form mercury vapors, and heating the slurry in a closed container to a temperature of at least about 50.degree. C. to produce vapors of the solvent and withdrawing vapors including solvent and mercury-containing vapors from the closed container, then separating mercury from the vapors withdrawn.

Akers, David J. (Indiana, PA); Ekechukwu, Kenneth N. (Silver Spring, MD); Aluko, Mobolaji E. (Burtonsville, MD); Lebowitz, Howard E. (Mountain View, CA)

2000-01-01T23:59:59.000Z

96

Process for removing NO sub x emissions from combustion effluents  

SciTech Connect

This patent describes a method of removing NO from a stream of combustion products injecting an alkyl amine into the stream to effect reduction of NO to N{sub 2}, wherein the alkyl amine comprises methyl amine, and the process is conducted at a temperature within the range of 350{degrees} C at a molar ratio to amine within the range of about 0.2 to 2.0 without a NO reduction catalyst, the method being capable of at least about a 50% conversion of NO in the combustion products to N{sub 2}.

Ham, D.O.; Moniz, G.A.; Gouveia, M.J.

1992-06-09T23:59:59.000Z

97

Biological removal of metal ions from aqueous process streams  

SciTech Connect

Aqueous waste streams from nuclear fuel processing operations may contain trace quantities of heavy metals such as uranium. Conventional chemical and physical treatment may be ineffective or very expensive when uranium concentrations in the range of 10 to 100 g/m/sup 3/ must be reduced to 1 g/m/sup 3/ or less. The ability of some microorganisms to adsorb or complex dissolved heavy metals offers an alternative treatment method. Uranium uptake by Saccharomyces cerevisiae NRRL Y-2574 and a strain of Pseudomonas aeruginosa was examined to identify factors which might affect a process for the removal of uranium from wastewater streams. At uranium concentrations in the range of 10 to 500 g/m/sup 3/, where the binding capacity of the biomass was not exceeded, temperature, pH, and initial uranium concentration were found to influence the rate of uranium uptake, but not the soluble uranium concentration at equilibrium. 6 figs.

Shumate, S.E. II; Strandberg, G.W.; Parrott, J.R. Jr.

1978-01-01T23:59:59.000Z

98

Actinide halide complexes  

DOE Patents (OSTI)

A compound is described of the formula MX[sub n]L[sub m] wherein M is a metal atom selected from the group consisting of thorium, plutonium, neptunium or americium, X is a halide atom, n is an integer selected from the group of three or four, L is a coordinating ligand selected from the group consisting of aprotic Lewis bases having an oxygen-, nitrogen-, sulfur-, or phosphorus-donor, and m is an integer selected from the group of three or four for monodentate ligands or is the integer two for bidentate ligands, where the sum of n+m equals seven or eight for monodentate ligands or five or six for bidentate ligands. A compound of the formula MX[sub n] wherein M, X, and n are as previously defined, and a process of preparing such actinide metal compounds are described including admixing the actinide metal in an aprotic Lewis base as a coordinating solvent in the presence of a halogen-containing oxidant.

Avens, L.R.; Zwick, B.D.; Sattelberger, A.P.; Clark, D.L.; Watkin, J.G.

1992-11-24T23:59:59.000Z

99

Actinide halide complexes  

DOE Patents (OSTI)

A compound of the formula MX.sub.n L.sub.m wherein M is a metal atom selected from the group consisting of thorium, plutonium, neptunium or americium, X is a halide atom, n is an integer selected from the group of three or four, L is a coordinating ligand selected from the group consisting of aprotic Lewis bases having an oxygen-, nitrogen-, sulfur-, or phosphorus-donor, and m is an integer selected from the group of three or four for monodentate ligands or is the integer two for bidentate ligands, where the sum of n+m equals seven or eight for monodentate ligands or five or six for bidentate ligands, a compound of the formula MX.sub.n wherein M, X, and n are as previously defined, and a process of preparing such actinide metal compounds including admixing the actinide metal in an aprotic Lewis base as a coordinating solvent in the presence of a halogen-containing oxidant, are provided.

Avens, Larry R. (Los Alamos, NM); Zwick, Bill D. (Santa Fe, NM); Sattelberger, Alfred P. (Los Alamos, NM); Clark, David L. (Los Alamos, NM); Watkin, John G. (Los Alamos, NM)

1992-01-01T23:59:59.000Z

100

2011 Actinide Separations Conference  

NLE Websites -- All DOE Office Websites (Extended Search)

of actinide chemistry and will play an important role in the future of the nuclear fuel cycle, nuclear medicine, and nuclear nonproliferation activities. The conference also...

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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101

Actinide Spectroscopy Workshop  

E-Print Network (OSTI)

the actinide series: Plutonium and Americium." The programin the 5f States of Plutonium. ” Gerrit van der Laan ofStructure of Uranium and Plutonium Compounds. ” Changing

Tobin, J.G.; Shuh, D.K.

2004-01-01T23:59:59.000Z

102

RAPID SEPARATION METHOD FOR ACTINIDES IN EMERGENCY AIR FILTER SAMPLES  

SciTech Connect

A new rapid method for the determination of actinides and strontium in air filter samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used in emergency response situations. The actinides and strontium in air filter method utilizes a rapid acid digestion method and a streamlined column separation process with stacked TEVA, TRU and Sr Resin cartridges. Vacuum box technology and rapid flow rates are used to reduce analytical time. Alpha emitters are prepared using cerium fluoride microprecipitation for counting by alpha spectrometry. The purified {sup 90}Sr fractions are mounted directly on planchets and counted by gas flow proportional counting. The method showed high chemical recoveries and effective removal of interferences. This new procedure was applied to emergency air filter samples received in the NRIP Emergency Response exercise administered by the National Institute for Standards and Technology (NIST) in April, 2009. The actinide and {sup 90}Sr in air filter results were reported in {approx}4 hours with excellent quality.

Maxwell, S.; Noyes, G.; Culligan, B.

2010-02-03T23:59:59.000Z

103

PREPARATION OF ACTINIDE-ALUMINUM ALLOYS  

DOE Patents (OSTI)

BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

Moore, R.H.

1962-09-01T23:59:59.000Z

104

Synthesis of actinide nitrides, phosphides, sulfides and oxides  

DOE Patents (OSTI)

This invention is comprised of a process of preparing an actinide compound of the formula An{sub x}Z{sub y} wherein An is an actinide metal atom selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, x is selected from the group consisting of one, two or three, Z is a main group element atom selected from the group consisting of nitrogen, phosphorus, oxygen and sulfur and y is selected from the group consisting of one, two, three or four, by admixing an actinide organometallic precursor wherein said actinide is selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, a suitable solvent and a protic Lewis base selected from the group consisting of ammonia, phosphine, hydrogen sulfide and water, at temperatures and for time sufficient to form an intermediate actinide complex, heating said intermediate actinide complex at temperatures and for time sufficient to form the actinide compound, and a process of depositing a thin film of such an actinide compound, e.g., uranium mononitride, by subliming an actinide organometallic precursor, e.g., a uranium amide precursor, in the presence of an effective amount of a protic Lewis base, e.g., ammonia, within a reactor at temperatures and for time sufficient to form a thin film of the actinide compound, are disclosed.

Van Der Sluys, W.G.; Burns, C.J.; Smith, D.C.

1991-04-02T23:59:59.000Z

105

Synthesis of actinide nitrides, phosphides, sulfides and oxides  

DOE Green Energy (OSTI)

A process of preparing an actinide compound of the formula An.sub.x Z.sub.y wherein An is an actinide metal atom selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, x is selected from the group consisting of one, two or three, Z is a main group element atom selected from the group consisting of nitrogen, phosphorus, oxygen and sulfur and y is selected from the group consisting of one, two, three or four, by admixing an actinide organometallic precursor wherein said actinide is selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, a suitable solvent and a protic Lewis base selected from the group consisting of ammonia, phosphine, hydrogen sulfide and water, at temperatures and for time sufficient to form an intermediate actinide complex, heating said intermediate actinide complex at temperatures and for time sufficient to form the actinide compound, and a process of depositing a thin film of such an actinide compound, e.g., uranium mononitride, by subliming an actinide organometallic precursor, e.g., a uranium amide precursor, in the presence of an effectgive amount of a protic Lewis base, e.g., ammonia, within a reactor at temperatures and for time sufficient to form a thin film of the actinide compound, are disclosed.

Van Der Sluys, William G. (Missoula, MT); Burns, Carol J. (Los Alamos, NM); Smith, David C. (Los Alamos, NM)

1992-01-01T23:59:59.000Z

106

A worldwide perspective on actinide burning  

SciTech Connect

Worldwide interest has been evident over the past few years in reexamining the merits of recovering the actinides from spent light-water reactor (LWR) fuel and transmuting them in fast reactors to reduce hazards in geologic repositories. This paper will summarize some of the recent activities in this field. Several countries are embarked on programs of reprocessing and vitrification of present wastes, from which removal of the actinides is largely precluded. The United States is assessing the ideas related to the fast reactor program and the potential application to defense wastes. 18 refs., 2 figs.

Burch, W.D.

1991-01-01T23:59:59.000Z

107

Overview of actinide chemistry in the WIPP  

Science Conference Proceedings (OSTI)

The year 2009 celebrates 10 years of safe operations at the Waste Isolation Pilot Plant (WIPP), the only nuclear waste repository designated to dispose defense-related transuranic (TRU) waste in the United States. Many elements contributed to the success of this one-of-the-kind facility. One of the most important of these is the chemistry of the actinides under WIPP repository conditions. A reliable understanding of the potential release of actinides from the site to the accessible environment is important to the WIPP performance assessment (PA). The environmental chemistry of the major actinides disposed at the WIPP continues to be investigated as part of the ongoing recertification efforts of the WIPP project. This presentation provides an overview of the actinide chemistry for the WIPP repository conditions. The WIPP is a salt-based repository; therefore, the inflow of brine into the repository is minimized, due to the natural tendency of excavated salt to re-seal. Reducing anoxic conditions are expected in WIPP because of microbial activity and metal corrosion processes that consume the oxygen initially present. Should brine be introduced through an intrusion scenario, these same processes will re-establish reducing conditions. In the case of an intrusion scenario involving brine, the solubilization of actinides in brine is considered as a potential source of release to the accessible environment. The following key factors establish the concentrations of dissolved actinides under subsurface conditions: (1) Redox chemistry - The solubility of reduced actinides (III and IV oxidation states) is known to be significantly lower than the oxidized forms (V and/or VI oxidation states). In this context, the reducing conditions in the WIPP and the strong coupling of the chemistry for reduced metals and microbiological processes with actinides are important. (2) Complexation - For the anoxic, reducing and mildly basic brine systems in the WIPP, the most important inorganic complexants are expected to be carbonate/bicarbonate and hydroxide. There are also organic complexants in TRU waste with the potential to strongly influence actinide solubility. (3) Intrinsic and pseudo-actinide colloid formation - Many actinide species in their expected oxidation states tend to form colloids or strongly associate with non actinide colloids present (e.g., microbial, humic and organic). In this context, the relative importance of actinides, based on the TRU waste inventory, with respect to the potential release of actinides from the WIPP, is greater for plutonium and americium, and to less extent for uranium and thorium. The most important oxidation states for WIPP-relevant conditions are III and IV. We will present an update of the literature on WIPP-specific data, and a summary of the ongoing research related to actinide chemistry in the WIPP performed by the Los Alamos National Laboratory (LANL) Actinide Chemistry and Repository Science (ACRSP) team located in Carlsbad, NM [Reed 2007, Lucchini 2007, and Reed 2006].

Borkowski, Marian [Los Alamos National Laboratory; Lucchini, Jean - Francois [Los Alamos National Laboratory; Richmann, Michael K [Los Alamos National Laboratory; Reed, Donald T [Los Alamos National Laboratory; Khaing, Hnin [Los Alamos National Laboratory; Swanson, Juliet [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

108

Removal of heavy metal ions from oil shale beneficiation process water by ferrite process  

SciTech Connect

The ferrite process is an established technique for removing heavy metals from waste water. Because the process water resulting from oil shale beneficiation falls into the category of industrial waste water, it is anticipated that this process may turn out to be a potential viable treatment for oil shale beneficiation process water containing many heave metal ions. The process is chemoremedial because not only effluent water comply with quality standards, but harmful heavy metals are converted into a valuable, chemically stable by-product known as ferrite. These spinel ferrites have magnetic properties, and therefore can be use in applications such as magnetic marker, ferrofluid, microwave absorbing and scavenging material. Experimental results from this process are presented along with results of treatment technique such as sulfide precipitation.

Mehta, R.K.; Zhang, L.; Lamont, W.E.; Schultz, C.W. (Alabama Univ., University, AL (United States). Mineral Resources Inst.)

1991-01-01T23:59:59.000Z

109

Removal of heavy metal ions from oil shale beneficiation process water by ferrite process  

SciTech Connect

The ferrite process is an established technique for removing heavy metals from waste water. Because the process water resulting from oil shale beneficiation falls into the category of industrial waste water, it is anticipated that this process may turn out to be a potential viable treatment for oil shale beneficiation process water containing many heave metal ions. The process is chemoremedial because not only effluent water comply with quality standards, but harmful heavy metals are converted into a valuable, chemically stable by-product known as ferrite. These spinel ferrites have magnetic properties, and therefore can be use in applications such as magnetic marker, ferrofluid, microwave absorbing and scavenging material. Experimental results from this process are presented along with results of treatment technique such as sulfide precipitation.

Mehta, R.K.; Zhang, L.; Lamont, W.E.; Schultz, C.W. [Alabama Univ., University, AL (United States). Mineral Resources Inst.

1991-12-31T23:59:59.000Z

110

A process for off-gas particulate removal  

DOE Patents (OSTI)

This paper describes an off-gas system for the removal of radioactive particulates from a melter for the vitrification of radioactive wastes to form glass waste forms. A diagram is provided.

Carl, D.E.

1998-04-01T23:59:59.000Z

111

Improved method for extracting lanthanides and actinides from acid solutions  

DOE Patents (OSTI)

A process for the recovery of actinide and lanthanide values from aqueous acidic solutions uses a new series of neutral bi-functional extractants, the alkyl(phenyl)-N,N-dialkylcarbamoylmethylphosphine oxides. The process is suitable for the separation of actinide and lanthanide values from fission product values found together in high-level nuclear reprocessing waste solutions.

Horwitz, E.P.; Kalina, D.G.; Kaplan, L.; Mason, G.W.

1983-07-26T23:59:59.000Z

112

Method for extracting lanthanides and actinides from acid solutions  

SciTech Connect

A process for the recovery of actinide and lanthanide values from aqueous acidic solutions with an organic extractant is patented. The process is suitable for the separation of actinide and lanthanide values from fission product values found together in high-level nuclear reprocessing waste solutions.

Horwitz, E.P.; Kalina, D.G.; Kaplan, L.; Mason, G.W.

1985-10-22T23:59:59.000Z

113

Advanced Extraction Methods for Actinide/Lanthanide Separations  

SciTech Connect

The separation of An(III) ions from chemically similar Ln(III) ions is perhaps one of the most difficult problems encountered during the processing of nuclear waste. In the 3+ oxidation states, the metal ions have an identical charge and roughly the same ionic radius. They differ strictly in the relative energies of their f- and d-orbitals, and to separate these metal ions, ligands will need to be developed that take advantage of this small but important distinction. The extraction of uranium and plutonium from nitric acid solution can be performed quantitatively by the extraction with the TBP (tributyl phosphate). Commercially, this process has found wide use in the PUREX (plutonium uranium extraction) reprocessing method. The TRUEX (transuranium extraction) process is further used to coextract the trivalent lanthanides and actinides ions from HLLW generated during PUREX extraction. This method uses CMPO [(N, N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide] intermixed with TBP as a synergistic agent. However, the final separation of trivalent actinides from trivalent lanthanides still remains a challenging task. In TRUEX nitric acid solution, the Am(III) ion is coordinated by three CMPO molecules and three nitrate anions. Taking inspiration from this data and previous work with calix[4]arene systems, researchers on this project have developed a C3-symmetric tris-CMPO ligand system using a triphenoxymethane platform as a base. The triphenoxymethane ligand systems have many advantages for the preparation of complex ligand systems. The compounds are very easy to prepare. The steric and solubility properties can be tuned through an extreme range by the inclusion of different alkoxy and alkyl groups such as methyoxy, ethoxy, t-butoxy, methyl, octyl, t-pentyl, or even t-pentyl at the ortho- and para-positions of the aryl rings. The triphenoxymethane ligand system shows promise as an improved extractant for both tetravalent and trivalent actinide recoveries form high level liquid wastes and a general actinide clean-up procedure. The selectivity of the standard extractant for tetravalent actinides, (N,N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide (CMPO), was markedly improved by the attachment of three CMPO-like functions onto a triphenoxymethane platform, and a ligand that is both highly selective and effective for An(IV) ions was isolated. A 10 fold excess of ligand will remove virtually all of the 4+ actinides from the acidic layer without extracting appreciable quantities of An(III) and Ln(III) unlike simple CMPO ligands. Inspired by the success of the DIAMEX industrial process for extractions, three new tripodal chelates bearing three diglycolamide and thiodiglycolamide units precisely arranged on a triphenoxymethane platform have been synthesized for an highly efficient extraction of trivalent f-element cations from nitric acid media. A single equivalent of ligand will remove 80% of the Ln(III) ion from the acidic layer since the ligand is perfectly suited to accommodate the tricapped trigonal prismatic geometry preferred by the metal center. The ligand is perhaps the most efficient binder available for the heavier lanthanides and due to this unique attribute, the extraction event can be easily followed by 1H NMR spectroscopy confirming the formation of a TPP complex. The most lipophilic di-n-butyl tris-diglycolamide was found to be a significantly weaker extractant in comparison to the di-isopropyl analogs. The tris-thiodiglycolamide derivative proved to be an ineffective chelate for f-elements and demonstrated the importance of the etheric oxygens in the metal binding. The results presented herein clearly demonstrate a cooperative action of these three ligating groups within a single molecule, confirmed by composition and structure of the extracted complexes, and since actinides prefer to have high coordination numbers, the ligands should be particularly adept at binding with three arms. The use of such an extractant permits the extraction of metal ions form highly acidic environment through the ability

Scott, M.J.

2005-12-01T23:59:59.000Z

114

Fluidized bed gasification ash reduction and removal process  

DOE Patents (OSTI)

In a fluidized bed gasification system an ash removal system to reduce the particulate ash to a maximum size or smaller, allow the ash to cool to a temperature lower than the gasifier and remove the ash from the gasifier system. The system consists of a crusher, a container containing level probes and a means for controlling the rotational speed of the crusher based on the level of ash within the container.

Schenone, Carl E. (Madison, PA); Rosinski, Joseph (Vanderbilt, PA)

1984-12-04T23:59:59.000Z

115

Improved Antimony Removal Using a Chemical Treatment and Microfiltration Process  

Science Conference Proceedings (OSTI)

Antimony removal can be a challenge because the species can exist in a number of valence states, in both soluble and insoluble forms. This report summarizes a test program conducted at Duke Power Company's Oconee plant, directed at removing antimony isotopes from the liquid radwaste stream. Treatments investigated included pH adjustment, use of oxidizing and reducing agents, application of seed materials, and addition of polyelectrolytes -- all combined with crossflow filtration. The report provides the ...

1998-06-30T23:59:59.000Z

116

Enzymatic Enhancement of Water Removal In the Dry Grind Corn to Ethanol Process.  

E-Print Network (OSTI)

??The removal of water from coproducts in the fuel ethanol process requires a significant energy input. The drying of the coproducts is responsible for as… (more)

Thomas, Ana Beatriz

2009-01-01T23:59:59.000Z

117

Thief process for the removal of mercury from flue gas  

DOE Patents (OSTI)

A system and method for removing mercury from the flue gas of a coal-fired power plant is described. Mercury removal is by adsorption onto a thermally activated sorbent produced in-situ at the power plant. To obtain the thermally activated sorbent, a lance (thief) is inserted into a location within the combustion zone of the combustion chamber and extracts a mixture of semi-combusted coal and gas. The semi-combusted coal has adsorptive properties suitable for the removal of elemental and oxidized mercury. The mixture of semi-combusted coal and gas is separated into a stream of gas and semi-combusted coal that has been converted to a stream of thermally activated sorbent. The separated stream of gas is recycled to the combustion chamber. The thermally activated sorbent is injected into the duct work of the power plant at a location downstream from the exit port of the combustion chamber. Mercury within the flue gas contacts and adsorbs onto the thermally activated sorbent. The sorbent-mercury combination is removed from the plant by a particulate collection system.

Pennline, Henry W. (Bethel Park, PA); Granite, Evan J. (Wexford, PA); Freeman, Mark C. (South Park Township, PA); Hargis, Richard A. (Canonsburg, PA); O' Dowd, William J. (Charleroi, PA)

2003-02-18T23:59:59.000Z

118

Process for removing polymer-forming impurities from naphtha fraction  

DOE Patents (OSTI)

Polymer precursor materials are vaporized without polymerization or are removed from a raw naphtha fraction by passing the raw naphtha to a vaporization zone and vaporizing the naphtha in the presence of a wash oil while stripping with hot hydrogen to prevent polymer deposits in the equipment. 2 figs.

Kowalczyk, D.C.; Bricklemyer, B.A.; Svoboda, J.J.

1983-12-27T23:59:59.000Z

119

Process for removing polymer-forming impurities from naphtha fraction  

DOE Patents (OSTI)

Polymer precursor materials are vaporized without polymerization or are removed from a raw naphtha fraction by passing the raw naphtha to a vaporization zone (24) and vaporizing the naphtha in the presence of a wash oil while stripping with hot hydrogen to prevent polymer deposits in the equipment.

Kowalczyk, Dennis C. (Pittsburgh, PA); Bricklemyer, Bruce A. (Avonmore, PA); Svoboda, Joseph J. (Pittsburgh, PA)

1983-01-01T23:59:59.000Z

120

NWChem and Actinide Chemistry  

NLE Websites -- All DOE Office Websites (Extended Search)

ACTINIDE CHEMISTRY MEETS COMPUTATION ACTINIDE CHEMISTRY MEETS COMPUTATION Capturing how contaminants migrate across groundwater-surface water inter- faces is a challenge that researchers at the Department of Energy's EMSL-the Environmental Molecular Sciences Laboratory-are rising to. This challenge, a top priority for waste cleanup efforts at the Hanford Site in Richland, Washington, and other parts of the DOE weapons complex, is being addressed using NWChem, a computational chemistry package developed at EMSL that is designed to run on high-performance parallel supercomputers, such as EMSL's Chinook. NWChem is enabling breakthrough discoveries in actinide behavior and chemistry, in part because it allows researchers to accurately model the dynamical formation, speciation, and redox chemistry of actinide complexes in realistic complex mo-

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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121

ACTINIDES-1981. ABSTRACTS  

E-Print Network (OSTI)

ACIDIC BOON TEMPERATURE MOLTEN SALT* R. De Waele, L. Heermanthe other actinides in molten salts » . This work describesAcinic Room Temperature Molten Salt R. De Waele, L. Heerman

Authors, Various

2010-01-01T23:59:59.000Z

122

Process for off-gas particulate removal and apparatus therefor  

DOE Patents (OSTI)

In the event of a breach in the off-gas line of a melter operation requiring closure of the line, a secondary vessel vent line is provided with a particulate collector utilizing atomization for removal of large particulates from the off-gas. The collector receives the gas containing particulates and directs a portion of the gas through outer and inner annular channels. The collector further receives a fluid, such as water, which is directed through the outer channel together with a second portion of the particulate-laden gas. The outer and inner channels have respective ring-like termination apertures concentrically disposed adjacent one another on the outer edge of the downstream side of the particulate collector. Each of the outer and inner channels curves outwardly away from the collector`s centerline in proceeding toward the downstream side of the collector. Gas flow in the outer channel maintains the fluid on the channel`s wall in the form of a ``wavy film,`` while the gas stream from the inner channel shears the fluid film as it exits the outer channel in reducing the fluid to small droplets. Droplets formed by the collector capture particulates in the gas stream by one of three mechanisms: impaction, interception or Brownian diffusion in removing the particulates. The particulate-laden droplets are removed from the fluid stream by a vessel vent condenser or mist eliminator. 4 figs.

Carl, D.E.

1997-10-21T23:59:59.000Z

123

Some Processing Options for the Removal of Critical Impurities from ...  

Science Conference Proceedings (OSTI)

Characterization of Indonesia Rare Earth Minerals and their Potential Processing Techniques · Characterization of Rare Earth Minerals with Field Emission ...

124

Bead and Process for Removing Dissolved Metal Contaminants  

DOE Patents (OSTI)

A bead is provided which comprises or consists essentially of activated carbon immobilized by crosslinked poly (carboxylic acid) binder, sodium silicate binder, or polyamine binder. The bead is effective to remove metal and other ionic contaminants from dilute aqueous solutions. A method of making metal-ion sorbing beads is provided, comprising combining activated carbon, and binder solution (preferably in a pin mixer where it is whipped), forming wet beads, and heating and drying the beads. The binder solution is preferably poly(acrylic acid) and glycerol dissolved in water and the wet beads formed from such binder solution are preferably heated and crosslinked in a convection oven.

Summers, Bobby L., Jr.; Bennett, Karen L.; Foster, Scott A.

2005-01-18T23:59:59.000Z

125

Nonaqueous actinide hydride dissolution and production of actinide $beta$- diketonates  

DOE Patents (OSTI)

Actinide beta-diketonate complex molecular compounds are produced by reacting a beta-diketone compound with a hydride of the actinide material in a mixture of carbon tetrachloride and methanol. (auth)

Crisler, L.R.

1975-11-11T23:59:59.000Z

126

Engineering-Scale Distillation of Cadmium for Actinide Recovery  

Science Conference Proceedings (OSTI)

During the recovery of actinide products from spent nuclear fuel, cadmium is separated from the actinide products by a distillation process. Distillation occurs in an induction-heated furnace called a cathode processor capable of processing kilogram quantities of cadmium. Operating parameters have been established for sufficient recovery of the cadmium based on mass balance and product purity. A cadmium distillation rate similar to previous investigators has also been determined. The development of cadmium distillation for spent fuel treatment enhances the capabilities for actinide recovery processes.

J.C. Price; D. Vaden; R.W. Benedict

2007-10-01T23:59:59.000Z

127

The ultra-high lime with aluminum process for removing chloride from recirculating cooling water  

E-Print Network (OSTI)

Chloride is a deleterious ionic species in cooling water systems because it is important in promoting corrosion. Chloride can be removed from cooling water by precipitation as calcium chloroaluminate using ultra-high lime with aluminum process (UHLA). The research program was conducted to study equilibrium characteristics and kinetics of chloride removal by UHLA process, study interactions between chloride and sulfate or silica, and develop a model for multicomponent removal by UHLA. Kinetics of chloride removal with UHLA was investigated. Chloride removal was found to be fast and therefore, removal kinetics should not be a limitation to applying the UHLA process. Equilibrium characteristics of chloride removal with UHLA were characterized. Good chloride removal was obtained at reasonable ranges of lime and aluminum doses. However, the stoichiometry of chloride removal with UHLA deviated from the theoretical stoichiometry of calcium chloroaluminate precipitation. Equilibrium modeling of experimental data and XRD analysis of precipitated solids indicated that this deviation was due to the formation of other solid phases such as tricalcium hydroxyaluminate and tetracalcium hydroxyaluminate. Effect of pH on chloride removal was characterized. Optimum pH for maximum chloride removal was pH 12 ± 0.2. Results of equilibrium experiments at different temperatures indicated that final chloride concentrations slightly increased when water temperature increased at temperatures below 40oC. However, at temperatures above 40oC, chloride concentration substantially increased with increasing water temperature. An equilibrium model was developed to describe chemical behavior of chloride removal from recycled cooling water using UHLA. Formation of a solid solution of calcium chloroaluminate, tricalcium hydroxyaluminate, and tetracalcium hydroxyaluminate was found to be the best mechanism to describe the chemical behavior of chloride removal with UHLA. Results of experiments that studied interactions between chloride and sulfate indicated that sulfate is preferentially removed over chloride. Final chloride concentration increased with increasing initial sulfate concentration. Silica was found to have only a small effect on chloride removal. The equilibrium model was modified in order to include sulfate and silica reactions along with chloride in UHLA process and it was able to accurately predict the chemical behavior of simultaneous removal of chloride, sulfate, and silica with UHLA.

Abdel-wahab, Ahmed Ibraheem Ali

2005-05-01T23:59:59.000Z

128

Development of an extraction process for removal of heteroatoms from coal liquids. Final report  

DOE Green Energy (OSTI)

The main goal of this contract was to develop an extraction process for upgrading coal liquids; and in doing so, to reduce the hydrogen requirement in downstream upgrading processes and to yield valuable byproducts. This goal was to be achieved by developing a novel carbon dioxide extraction process for heteroatom removal from coal-derived naphtha, diesel, and jet fuel. The research plan called for the optimization of three critical process variables using a statistically-designed experimental matrix. The commercial potential of the new process was to be evaluated by demonstrating quantitatively the effectiveness of heteroatom removal from three different feedstocks and by conducting a comparative economic analysis of alternate heteroatom removal technologies. Accomplishments are described for the following tasks: food procurement and analysis process variable screening studies; and process assessment.

Not Available

1994-04-01T23:59:59.000Z

129

Determining the removal effectiveness of flame retardants from drinking water treatment processes  

E-Print Network (OSTI)

Low concentrations of xenobiotic chemicals have recently become a concern in the surface water environment. The concern expands to drinking water treatment processes, and whether or not they remove these chemicals while ...

Lin, Joseph C. (Joseph Chris), 1981-

2004-01-01T23:59:59.000Z

130

Processes to remove acid forming gases from exhaust gases  

DOE Patents (OSTI)

The present invention relates to a process for reducing the concentration of NO in a gas, which process comprises: (A) contacting a gas sample containing NO with a gaseous oxidizing agent to oxidize the NO to NO[sub 2]; (B) contacting the gas sample of step (A) comprising NO[sub 2] with an aqueous reagent of bisulfite/sulfite and a compound selected from urea, sulfamic acid, hydrazinium ion, hydrazoic acid, nitroaniline, sulfanilamide, sulfanilic acid, mercaptopropanoic acid, mercaptosuccinic acid, cysteine or combinations thereof at between about 0 and 100 C at a pH of between about 1 and 7 for between about 0.01 and 60 sec; and (C) optionally contacting the reaction product of step (A) with conventional chemical reagents to reduce the concentrations of the organic products of the reaction in step (B) to environmentally acceptable levels. Urea or sulfamic acid are preferred, especially sulfamic acid, and step (C) is not necessary or performed. 16 figs.

Chang, S.G.

1994-09-20T23:59:59.000Z

131

Device for Detecting Actinides, Method for Detecting Actinides  

DOE Patents (OSTI)

A heavy metal detector is provided comprising a first molecule and a second molecule, whereby the first and second molecules interact in a predetermined manner; a first region on the first molecule adapted to interact with an actinide; and a second region on the second molecule adapted to interact with the actinide, whereby the interactions of the actinide with the regions effect the predetermined manner of interaction between the molecules.

Stevens, Fred J.; Wilkins-Stevens, Priscilla

1998-10-29T23:59:59.000Z

132

Method and system for the removal of oxides of nitrogen and sulfur from combustion processes  

DOE Patents (OSTI)

A process for removing oxide contaminants from combustion gas, and employing a solid electrolyte reactor, includes: (a) flowing the combustion gas into a zone containing a solid electrolyte and applying a voltage and at elevated temperature to thereby separate oxygen via the solid electrolyte, (b) removing oxygen from that zone in a first stream and removing hot effluent gas from that zone in a second stream, the effluent gas containing contaminant, (c) and pre-heating the combustion gas flowing to that zone by passing it in heat exchange relation with the hot effluent gas.

Walsh, John V. (Glendora, CA)

1987-12-15T23:59:59.000Z

133

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

SciTech Connect

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

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

2011-02-07T23:59:59.000Z

134

Heavy Actinides | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

actinides with the construction and commissioning of the High Flux Isotope Reactor (HFIR) and Radiochemical Engineering Development Center (REDC) in 1965 and 1966,...

135

DEVELOPMENT OF CHEMICAL REDUCTION AND AIR STRIPPING PROCESSES TO REMOVE MERCURY FROM WASTEWATER  

SciTech Connect

This study evaluates the removal of mercury from wastewater using chemical reduction and air stripping using a full-scale treatment system at the Savannah River Site. The existing water treatment system utilizes air stripping as the unit operation to remove organic compounds from groundwater that also contains mercury (C ~ 250 ng/L). The baseline air stripping process was ineffective in removing mercury and the water exceeded a proposed limit of 51 ng/L. To test an enhancement to the existing treatment modality a continuous dose of reducing agent was injected for 6-hours at the inlet of the air stripper. This action resulted in the chemical reduction of mercury to Hg(0), a species that is removable with the existing unit operation. During the injection period a 94% decrease in concentration was observed and the effluent satisfied proposed limits. The process was optimized over a 2-day period by sequentially evaluating dose rates ranging from 0.64X to 297X stoichiometry. A minimum dose of 16X stoichiometry was necessary to initiate the reduction reaction that facilitated the mercury removal. Competing electron acceptors likely inhibited the reaction at the lower 1 doses, which prevented removal by air stripping. These results indicate that chemical reduction coupled with air stripping can effectively treat large-volumes of water to emerging part per trillion regulatory standards for mercury.

Jackson, D.; Looney, B.; Craig, B.; Thompson, M.; Kmetz, T.

2013-07-10T23:59:59.000Z

136

A Heat Exchanger Process for Removal of H{sub2}S Gas  

SciTech Connect

A heat exchanger process has been developed for the removal of H{sub 2}S and other noncondensable gases from geothermal steam. The process utilizes a heat exchanger to condense water from geothermal steam while allowing H{sub 2}S and other noncondensable gases to pass through in the vapor phase. The condensed water is evaporated to form a clean steam from which over 90 percent of the H{sub 2}S and other noncondensable gases have been removed. Some of the important advantages of the heat exchanger process are shown in Table 1. The system can be located upstream of a power plant turbine which eliminates much of the potential for corrosion, as well as the requirement for removing H{sub 2}S from water collected in the main condenser. Since almost all noncondensables are removed, much less steam is needed for air ejector operation. The heat exchanger process is simple: it has no chemical addition requirements or sludge by-products and utilizes standard equipment found in many power plant applications. The regular power plant operators and maintenance crews can easily understand and run the system with minimal attention. Capital and operating costs are competitive with those for currently available H{sub 2}S-abatement technology, although significant economic advantages over downstream abatement processes may result due to the use of clean steam in the turbines.

Coury, Glenn E.; Babione, Robert A.; Gosik, Robert J.

1980-12-01T23:59:59.000Z

137

Rapid Separation Methods to Characterize Actinides and Metallic Impurities in Plutonium Scrap Materials at SRS  

SciTech Connect

The Nuclear Materials Stabilization and Storage Division at SRS plans to stabilize selected plutonium scrap residue materials for long term storage by dissolution processing and plans to stabilize other plutonium vault materials via high-temperature furnace processing. To support these nuclear material stabilization activities, the SRS Analytical Laboratories Department (ALD) will provide characterization of materials required prior to the dissolution or the high-firing of these materials. Lab renovations to install new analytical instrumentation are underway to support these activities that include glove boxes with simulated-process dissolution and high- pressure microwave dissolution capability. Inductively-coupled plasma atomic emission spectrometry (ICP-AES), inductively- coupled mass spectrometry (ICP-MS) and thermal-ionization mass spectrometry (TIMS) will be used to measure actinide isotopics and metallic impurities. New high-speed actinide separation methods have been developed that will be applied to isotopic characterization of nuclear materials by TIMS and ICP-MS to eliminate isobaric interferences between Pu-238 /U- 238 and Pu-241/Am-241. TEVA Resin, UTEVA Resin, and TRU Resin columns will be used with vacuum-assisted flow rates to minimize TIMS and ICP-MS sample turnaround times. For metallic impurity analysis, rapid column removal methods using UTEVA Resin, AGMP-1 anion resin and AG MP-50 cation resin have also been developed to remove plutonium and uranium matrix interferences prior to ICP-AES and ICP- MS measurements.

Maxwell, S.L. III [Westinghouse Savannah River Company, AIKEN, SC (United States); Jones, V.D.

1998-07-01T23:59:59.000Z

138

Process for removal of hydrogen halides or halogens from incinerator gas  

DOE Patents (OSTI)

A process for reducing the amount of halogens and halogen acids in high temperature combustion gases and through their removal, the formation of halogenated organics at lower temperatures, with the reduction being carried out electrochemically by contacting the combustion gas with the negative electrode of an electrochemical cell and with the halogen and/or halogen acid being recovered at the positive electrode.

Huang, Hann S. (Darien, IL); Sather, Norman F. (Naperville, IL)

1988-01-01T23:59:59.000Z

139

Process for removal of hydrogen halides or halogens from incinerator gas  

DOE Patents (OSTI)

A process for reducing the amount of halogens and halogen acids in high temperature combustion gas and through their removal, the formation of halogenated organics at lower temperatures, with the reduction being carried out electrochemically by contacting the combustion gas with the negative electrode of an electrochemical cell and with the halogen and/or halogen acid being recovered at the positive electrode.

Huang, H.S.; Sather, N.F.

1987-08-21T23:59:59.000Z

140

Development of Acetic Acid Removal Technology for the UREX+Process  

SciTech Connect

It is imperative that acetic acid is removed from a waste stream in the UREX+process so that nitric acid can be recycled and possible interference with downstreatm steps can be avoidec. Acetic acid arises from acetohydrozamic acid (AHA), and is used to suppress plutonium in the first step of the UREX+process. Later, it is hydrolyzed into hydroxyl amine nitrate and acetic acid. Many common separation technologies were examined, and solvent extraction was determined to be the best choice under process conditions. Solvents already used in the UREX+ process were then tested to determine if they would be sufficient for the removal of acetic acid. The tributyl phosphage (TBP)-dodecane diluent, used in both UREX and NPEX, was determined to be a solvent system that gave sufficient distribution coefficients for acetic acid in addition to a high separation factor from nitric acid.

Robert M. Counce; Jack S. Watson

2009-06-30T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Removal of strontium-90 from calcined wastes with the SREX process  

SciTech Connect

Experiments have been performed to formulate chemical procedures for the processing of radioactive dissolved calcine wastes for the removal of {sup 90}Sr with the SREX (Strontium Extraction) solvent. Batch contact solvent extraction experiments have been performed to yield a processing scheme which is highly efficient for the extraction of Sr, while remaining free from insoluble precipitate formation and third phase formation. The effect of various scrubbing and stripping techniques and elevated temperature have been evaluated. The results of the batch contact experimentation has formed the basis for a proposed processing flowsheet which is scheduled soon for demonstration in centrifugal contactors.

Wood, D.J.; Mann, N.R.; Tillotson, R.; Tullock, P.A.; Todd, T.A.

1997-12-31T23:59:59.000Z

142

Electrorecovery of actinides at room temperature  

Science Conference Proceedings (OSTI)

There are a large number of purification and processing operations involving actinide species that rely on high-temperature molten salts as the solvent medium. One such application is the electrorefining of impure actinide metals to provide high purity material for subsequent applications. There are some drawbacks to the electrodeposition of actinides in molten salts including relatively low yields, lack of accurate potential control, maintaining efficiency in a highly corrosive environment, and failed runs. With these issues in mind we have been investigating the electrodeposition of actinide metals, mainly uranium, from room temperature ionic liquids (RTILs) and relatively high-boiling organic solvents. The RTILs we have focused on are comprised of 1,3-dialkylimidazolium or quaternary ammonium cations and mainly the {sup -}N(SO{sub 2}CF{sub 3}){sub 2} anion [bis(trif1uoromethylsulfonyl)imide {equivalent_to} {sup -}NTf{sub 2}]. These materials represent a class of solvents that possess great potential for use in applications employing electrochemical procedures. In order to ascertain the feasibility of using RTILs for bulk electrodeposition of actinide metals our research team has been exploring the electron transfer behavior of simple coordination complexes of uranium dissolved in the RTIL solutions. More recently we have begun some fundamental electrochemical studies on the behavior of uranium and plutonium complexes in the organic solvents N-methylpyrrolidone (NMP) and dimethylsulfoxide (DMSO). Our most recent results concerning electrodeposition will be presented in this account. The electrochemical behavior of U(IV) and U(III) species in RTILs and the relatively low vapor pressure solvents NMP and DMSO is described. These studies have been ongoing in our laboratory to uncover conditions that will lead to the successful bulk electrodeposition of actinide metals at a working electrode surface at room temperature or slightly elevated temperatures. The RTILs we have focused on thus far are based on 1,3-dialkylimidazolium or quaternary ammonium cations and {sup -}N(SO{sub 2}CF{sub 3}){sub 2} anions. Our results from XPS studies of e1ectrooxidized uranium metal surfaces indicate that uranium metal reacts with the anion from the RTIL, most likely through an initial f1uoride abstraction, forming decomposition products that inhibit the bulk electrodeposition of uranium metal. Similar results were found when the organic solvents were used with TBA[B(C{sub 6}F{sub 5}){sub 4}] as the supporting electrolyte, although the voltammetric data of uranium ions in these solutions is more encouraging in relation to electrodeposition of uranium metal. Preliminary results on the voltammetric behavior and bulk electrodeposition of plutonium species are also presented.

Stoll, Michael E [Los Alamos National Laboratory; Oldham, Warren J [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory

2008-01-01T23:59:59.000Z

143

Recycle of LWR (Light Water Reactor) actinides to an IFR (Integral Fast Reactor)  

SciTech Connect

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

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

1991-01-01T23:59:59.000Z

144

Actinide Burning in CANDU Reactors  

Science Conference Proceedings (OSTI)

Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

Hyland, B.; Dyck, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

2007-07-01T23:59:59.000Z

145

Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries  

DOE Patents (OSTI)

A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper(II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the original organic compounds, is subsequently blended with high level radioactive sludge land transferred to a vitrification facility for processing into borosilicate glass for long-term storage. 2 figs., 3 tabs.

Doherty, J.P.; Marek, J.C.

1987-02-25T23:59:59.000Z

146

TECHNICAL AND OPERATING SUPPORT FOR PILOT DEMONSTRATION OF MORPHYSORB ACID GAS REMOVAL PROCESS  

Science Conference Proceedings (OSTI)

GTI and Krupp Uhde have been jointly developing advanced technology for removing high concentrations of acid gas from high-pressure natural gas for over a decade. This technology, the Morphysorb{reg_sign} process, based on N-formyl and N-acetyl morpholine mixtures, has now been tested in a large-scale facility and this paper presents preliminary results from acceptance testing at that facility. Earlier publications have discussed the bench-scale and pilot plant work that led up to this important milestone. The site was Duke Energy's new Kwoen sour gas upgrader near Chetwynd B.C., Canada. This facility has a nameplate capacity of 300 MMscfd of sour natural gas. The objective of the Morphysorb process at this site was to remove 33 MMscfd of acid gas (H{sub 2}S and CO{sub 2}) for reinjection downhole. This represents about half the acid gas present in the feed to the plant. In so doing, proportionately more of the plant ''sales'' gas, which is sent for final processing at the nearby Pine River plant, can be sent down the line without coming up against the sulfur removal capacity limits of Pine River plant, than could with other solvents that were evaluated. Other benefits include less loss of methane downhole with the rejected acid gas and lower circulation and recycle compression horsepower than with competitive solvents. On the downside, the process is expected to have higher solvent vaporization losses than competitive solvents, but this is a comparatively minor drawback when weighed against the value of the benefits. These benefits (and drawbacks) were developed into quantitative ''acceptance'' criteria, which will determine if the solvent will continue to be used at the site and for award of monetary bonuses to the process developer (GTI).

Nagaraju Palla; Dennis Leppin

2003-09-30T23:59:59.000Z

147

TECHNICAL AND OPERATING SUPPORT FOR PILOT DEMONSTRATION OF MORPHYSORB ACID GAS REMOVAL PROCESS  

Science Conference Proceedings (OSTI)

GTI and Krupp Uhde have been jointly developing advanced technology for removing high concentrations of acid gas from high-pressure natural gas for over a decade. This technology, the Morphysorb{reg_sign} process, based on N-formyl and N-acetyl morpholine mixtures, has now been tested in a large-scale facility and this paper presents preliminary results from acceptance testing at that facility. Earlier publications have discussed the bench-scale and pilot plant work that led up to this important milestone. The site was Duke Energy's new Kwoen sour gas upgrader near Chetwynd B.C., Canada. This facility has a nameplate capacity of 300 MMscfd of sour natural gas. The objective of the Morphysorb process at this site was to remove 33 MMscfd of acid gas (H{sub 2}S and CO{sub 2}) for reinjection downhole. This represents about half the acid gas present in the feed to the plant. In so doing, proportionately more of the plant ''sales'' gas, which is sent for final processing at the nearby Pine River plant, can be sent down the line without coming up against the sulfur removal capacity limits of Pine River plant, than could with other solvents that were evaluated. Other benefits include less loss of methane downhole with the rejected acid gas and lower circulation and recycle compression horsepower than with competitive solvents. On the downside, the process is expected to have higher solvent vaporization losses than competitive solvents, but this is a comparatively minor drawback when weighed against the value of the benefits. These benefits (and drawbacks) were developed into quantitative ''acceptance'' criteria, which will determine if the solvent will continue to be used at the site and for award of monetary bonuses to the process developer (GTI).

Nagaraju Palla; Dennis Leppin

2003-06-30T23:59:59.000Z

148

Process for removing sulfur from sulfur-containing gases: high calcium fly-ash  

DOE Patents (OSTI)

The present disclosure relates to improved processes for treating hot sulfur-containing flue gas to remove sulfur therefrom. Processes in accordance with the present invention include preparing an aqueous slurry composed of a calcium alkali source and a source of reactive silica and/or alumina, heating the slurry to above-ambient temperatures for a period of time in order to facilitate the formation of sulfur-absorbing calcium silicates or aluminates, and treating the gas with the heat-treated slurry components. Examples disclosed herein demonstrate the utility of these processes in achieving improved sulfur-absorbing capabilities. Additionally, disclosure is provided which illustrates preferred configurations for employing the present processes both as a dry sorbent injection and for use in conjunction with a spray dryer and/or bagfilter. Retrofit application to existing systems is also addressed.

Rochelle, Gary T. (Austin, TX); Chang, John C. S. (Cary, NC)

1991-01-01T23:59:59.000Z

149

Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream  

SciTech Connect

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, {sup 203}Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste simulant and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl{sub 2} from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO{sub 3} and 0.077 with 0.25 M Na{sub 2}CO{sub 3}. An experimental flowsheet was designed from the batch contact tests and tested counter-currently using 5.5 cm centrifugal contactors. Results from the counter-current test show that mercury can be removed from the acidic mixed SBW simulant and recovered separately from the actinides.

Brewer, K.N.; Herbst, R.S.; Tranter, T.J. [and others

1995-12-01T23:59:59.000Z

150

Method for extracting lanthanides and actinides from acid solutions  

SciTech Connect

A process for the recovery of actinide and lanthanide values from aqueous acidic solutions with an organic extractant having the formula: ##STR1## where .phi. is phenyl, R.sup.1 is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R.sup.2 is an alkyl containing from 3 to 6 carbon atoms. The process is suitable for the separation of actinide and lanthanide values from fission product values found together in high level nuclear reprocessing waste solutions.

Horwitz, E. Philip (Naperville, IL); Kalina, Dale G. (Naperville, IL); Kaplan, Louis (Lombard, IL); Mason, George W. (Clarendon Hills, IL)

1985-01-01T23:59:59.000Z

151

Analysis of hypochlorite process for removal of hydrogen sulfide from geothermal gases  

SciTech Connect

Sodium hypochlorite reacts readily with hydrogen sulfide to convert the sulfide ion into free sulfur in a neutral or acid solution and to the sulfate ion in an alkaline solution. Sodium hypochlorite can be generated on site by processing geothermal brine in electrolytic cells. An investigation to determine if this reaction could be economically used to remove hydrogen sulfide from geothermal noncondensible gases is reported. Two processes, the LO-CAT Process and the Stretford Process, were selected for comparison with the hypochlorite process. Three geothermal reservoirs were considered for evaluation: Niland KGRA, Baca KGRA, and The Geysers KGRA. Because of the wide variation in the amount of hydrogen sulfide present at The Geysers, two different gas analyses were considered for treatment. Plants were designed to process the effluent noncondensible gases from a 10 MW/sub e/ geothermal power plant. The effluent gas from each plant was to contain a maximum hydrogen sulfide concentration of 35 ppb. Capital costs were estimated for each of the processes at each of the four sites selected. Operating costs were also calculated for each of the processes at each of the sites. The results of these studies are shown.

1980-04-01T23:59:59.000Z

152

Thermal removal of mercury in spent powdered activated carbon from TOXECON process  

SciTech Connect

This research developed and demonstrated a technology to liberate Hg adsorbed onto powdered activated carbon (PAC) by the TOXECON process using pilot-scale high temperature air slide (HTAS) and bench-scale thermogravimetric analyzer (TGA). The HTAS removed 65, 83, and 92% of Hg captured with PAC when ran at 900{sup o}F, 1,000{sup o}F, and 1,200 {sup o}F, respectively, while the TGA removed 46 and 100% of Hg at 800 {sup o}F and 900{sup o}F, respectively. However, addition of CuO-Fe{sub 2}O{sub 3} mixture and CuCl catalysts enhanced Hg removal and PAC regeneration at lower temperatures. CuO-Fe{sub 2}O{sub 3} mixture performed better than CuCl in PAC regeneration. Scanning electron microscopy images and energy dispersive X-ray analysis show no change in PAC particle aggregation or chemical composition. Thermally treated sorbents had higher surface area and pore volume than the untreated samples indicating regeneration. The optimum temperature for PAC regeneration in the HTAS was 1,000{sup o}F. At this temperature, the regenerated sorbent had sufficient adsorption capacity similar to its virgin counterpart at 33.9% loss on ignition. Consequently, the regenerated PAC may be recycled back into the system by blending it with virgin PAC.

Okwadha, G.D.O.; Li, J.; Ramme, B.; Kollakowsky, D.; Michaud, D. [University of Wisconsin, Milwaukee, WI (United States)

2009-10-15T23:59:59.000Z

153

Removal of plutonium and uranium from process streams using ultrafiltration membranes  

SciTech Connect

A series of experiments using hollow fiber ultrafiltration modules was run on various Mound Laboratory waste streams contaminated with /sup 238/Pu, /sup 239/Pu, and /sup 233/U. These modules had various molecular weight cut-offs ranging from 2000 to 80,000. The types of waste solution studied consisted of waste water from the ''hot'' laundry, decontamination water from the Plutonium Processing (PP) Building, and influent to the Waste Disposal (WD) Building. These experiments have shown that the ability to remove radioactivity is a function of the contents of the waste stream. This is due to the fact that the radioactivity in the waste water is in various forms (ionic, polymeric, colloidal, and adsorbed onto suspended solids). Removal of suspended or colloidal material was very high, while removal of ionic material was very low. The best case proved to be the laundry waste water which yielded a rejection of radioactivity up to 99.8%, with a product concentration of <0.1 dis/min/ml. The worst case was decontamination water which yielded a rejection of radioactivity of 85 to 88% with a product concentration of 166 to 229 dis/min/ml (initial feed was 1440 dis/min/ml). Typical WD influent showed a rejection of radioactivity of 90 to 98% and a product concentration of from 7 to 100 dis/min/ml, depending upon initial concentration and the nature of the waste stream.

Roberts, R.C.; Koenst, J.W.

1977-01-01T23:59:59.000Z

154

Process for removing thorium and recovering vanadium from titanium chlorinator waste  

DOE Patents (OSTI)

A process for removal of thorium from titanium chlorinator waste comprising: (a) leaching an anhydrous titanium chlorinator waste in water or dilute hydrochloric acid solution and filtering to separate insoluble minerals and coke fractions from soluble metal chlorides; (b) beneficiating the insoluble fractions from step (a) on shaking tables to recover recyclable or otherwise useful TiO.sub.2 minerals and coke; and (c) treating filtrate from step (a) with reagents to precipitate and remove thorium by filtration along with acid metals of Ti, Zr, Nb, and Ta by the addition of the filtrate (a), a base and a precipitant to a boiling slurry of reaction products (d); treating filtrate from step (c) with reagents to precipitate and recover an iron vanadate product by the addition of the filtrate (c), a base and an oxidizing agent to a boiling slurry of reaction products; and (e) treating filtrate from step (d) to remove any remaining cations except Na by addition of Na.sub.2 CO.sub.3 and boiling.

Olsen, Richard S. (Albany, OR); Banks, John T. (Corvallis, OR)

1996-01-01T23:59:59.000Z

155

Method for removing volatile components from a ceramic article, and related processes  

DOE Patents (OSTI)

A method of removing substantially all of the volatile component in a green, volatile-containing ceramic article is disclosed. The method comprises freezing the ceramic article; and then subjecting the frozen article to a vacuum for a sufficient time to freeze-dry the article. Frequently, the article is heated while being freeze-dried. Use of this method efficiently reduces the propensity for any warpage of the article. The article is often formed from a ceramic slurry in a gel-casting process. A method for fabricating a ceramic core used in investment casting is also described.

Klug, Frederic Joseph (Schenectady, NY); DeCarr, Sylvia Marie (Waterford, NY)

2002-01-01T23:59:59.000Z

156

Cyclic process for producing methane in a tubular reactor with effective heat removal  

DOE Patents (OSTI)

Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

Frost, Albert C. (Congers, NY); Yang, Chang-Lee (Spring Valley, NY)

1986-01-01T23:59:59.000Z

157

Cyclic process for producing methane from carbon monoxide with heat removal  

DOE Patents (OSTI)

Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

Frost, Albert C. (Congers, NY); Yang, Chang-lee (Spring Valley, NY)

1982-01-01T23:59:59.000Z

158

On the Development of a Distillation Process for the Electrometallurgical Treatment of Irradiated Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

B.R. Westphal; K.C. Marsden; J.C. Price; D.V. Laug

2008-04-01T23:59:59.000Z

159

TECHNICAL AND OPERATING SUPPORT FOR PILOT DEMONSTRATION OF MORPHYSORB ACID GAS REMOVAL PROCESS  

SciTech Connect

Over the past 14 years, the Gas Technology Institute and jointly with Uhde since 1997 developing Morphysorb{reg_sign} a new physical solvent-based acid gas removal process. Based on extensive laboratory, bench, pilot-plant scale experiments and computer simulations, DEGT Gas Transmission Company, Canada (DEGT) has chosen the process for use at its Kwoen processing facility near Chetwynd, British Columbia, Canada as the first commercial application for the Morphysorb process. DOE co-funded the development of the Morphysorb process in various stages of development. DOE funded the production of this report to ensure that the results of the work would be readily available to potential users of the process in the United States. The Kwoen Plant is designed to process 300 MMscfd of raw natural gas at 1,080-psia pressure. The sour natural gas contains 20 to 25 percent H{sub 2}S and CO{sub 2}. The plant reduces the acid gas content by about 50% and injects the removed H{sub 2}S and CO{sub 2} into an injection well. The Kwoen plant has been operating since August 2002. Morphysorb{reg_sign} is a physical solvent-based process used for the bulk removal of CO{sub 2} and/or H{sub 2}S from natural gas and other gaseous streams. The solvent consists of N-Formyl morpholine and other morpholine derivatives. This process is particularly effective for high-pressure and high acid-gas applications and offers substantial savings in investment and operating cost compared to competitive physical solvent-based processes. GTI and DEGT first entered into an agreement in 2002 to test the Morphysorb process at their Kwoen Gas Treating Plant in northern BC. The process is operating successfully without any solvent related problems and has between DEGTC and GTI. As of December 2003, about 90 Bcf of sour gas was processed. Of this about 8 Bcf of acid gas containing mainly H{sub 2}S and CO{sub 2} was injected back into the depleted reservoir and 82 Bcf sent for further processing at DEGTC's Pine River Plant. This report discusses the operational performance at Kwoen plant during the performance test as well as the solvent performance since the plant started up. The Morphysorb performance is assessed by Duke Energy according to five metrics: acid gas pickup, recycle gas flow, total hydrocarbon loss in acid gas stream, Morphysorb solvent losses and foaming related problems. Plant data over a period of one year show that the Morphysorb solvent has performed extremely well in four out of five of these categories. The fifth metric, Morphysorb solvent loss, is being evaluated over a longer-term period in order to accurately assess it. However, the preliminary indications based on makeup solvent used to date are that solvent losses will also be within expectations. The analysis of the solvent samples indicates that the solvent is very stable and did not show any sign of degradation. The operability of the solvent is good and no foaming related problems have been encountered. According to plant operators the Morphysorb unit runs smoothly and requires no special attention.

Nagaraju Palla; Dennis Leppin

2004-02-01T23:59:59.000Z

160

CMP flowsheet development for the separation of actinides from ICPP sodium-bearing waste using centrifugal contactors  

Science Conference Proceedings (OSTI)

Previous results of lab-scale batch contacts with sodium-bearing waste (SBW) simulant suggested a potential flowsheet for partitioning actinides using solvent extraction with dihexyl-N,N-diethylcarbamoylmethyl phosphonate (DHDECMP or simply CMP) as the extractant. The suggested baseline flowsheet includes: an extraction section to remove actinides from liquid SBW into the CMP solvent (0.75 M CMP, 1.0 M TBP in Isopar-L{reg_sign}); a thermally unstable complexant (TUCS) strip section to back-extract actinides; a sodium carbonate wash section for solvent cleanup; and a dilute HNO{sub 3} rinse section to re-acidify the solvent. The purpose of these studies was to test and develop a baseline CMP flowsheet for Idaho Chemical Processing Plant (ICPP) SBW under continuous, countercurrent conditions using centrifugal contactors. This flowsheet was tested in two experiments using the Centrifugal Contactor Mockup which consists of sixteen stages of 5.5 cm diameter centrifugal contactors (procured from Oak Ridge National Laboratory). All testing was performed using non-radioactive SBW simulant. Potential flowsheets were evaluated with regard to the behavior of the non-radioactive components potentially extracted by the CMP solvent. Specifically, the behavior of the matrix components, including Fe, Hg, and Zr, was studied. In addition, Nd was added to the SBW simulant as a surrogate for {sup 241}Am. In general, the behavior of the individual components closely paralleled that anticipated from batch testing. Based on the assumption that the behavior of Am will be very similar to the behavior of the Nd surrogate, eight extraction stages are more than sufficient to reduce the actinide content in the SBW to levels well below the NRC Class A LLW criteria of 10 nCi/g. Very little Fe or Zr were extracted from the SBW simulant, resulting in only 1% of the Fe and 4% of the Zr exiting in the high-activity waste (HAW) fraction.

Law, J.D.; Herbst, R.S.; Rodriguez, A.M.

1995-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Assisted thermal stripping (ATS) for removal of PCBs from contaminated soils. Design of experiments modeling of the ATS process  

Science Conference Proceedings (OSTI)

In a companion report, the Assisted Thermal Stripping (ATS) process for enhanced removal of PCBs from PCB-contaminated soil is described. In studies directed toward achieving residual PCB levels of {le}2 ppm, it was found that four factors were particularly important -- (1) process temperature; (2) process time; (3) the amount of additive (for enhancing the removal of PCBs); and (4) steam flow rate. In order to optimize the ATS process, it was deemed crucial to ascertain the relative effect exerted by each of those process factors and the reproducibility of the process. To accomplish that, we have relied on the technique {open_quotes}Design of Experiments{close_quotes} (DOE) to mathematically model the ATS process. After considering the findings from our previous investigations, it was decided to employ formic acid as the additive for enhancing the removal of PCBs.

Krabbenhoft, H.O.; Webb, J.L.; Gascoyne, D.G. [GE Corporate Research & Development, Schenectady, NY (United States); Cawse, J.N. [GE Plastics, Pittsfield, MA (United States)

1996-12-31T23:59:59.000Z

162

H[sub 2]S in EOR--1: Gas processing for CO[sub 2] EOR involves sulfur removal  

SciTech Connect

A design study for a new West Texas gas processing plant for a CO[sub 2] EOR project provides for installation of H[sub 2]S removal processes to be delayed for 3 years after completion of the plant. During this delay, a more precise produced gas composition will be obtained so that the process equipment for removing H[sub 2]S can be properly selected and sized to handle the gas stream that at the peak will reach about 30 MMscfd. The new plant's processing components include inlet separation, sulfur removal and recovery, compression, dehydration, and NGL recovery. The new plant will be capable of processing CO[sub 2]-contaminated associated gas, recovering valuable propane-plus NGLs,a nd producing a miscible CO[sub 2] for reinjection. The first in a series of two articles details the process and configuration options. The concluding part will discuss in greater detail the sulfur recovery alternatives.

Johnson, J.E. (Pritchard Corp., Overland Park, KS (United States)); Tzap, S.J.; Kelley, R.E. (Raytheon Engineers and Constructors, Denver, CO (United States)); Laczko, L.P. (OXY USA Inc., Midland, TX (United States))

1993-11-15T23:59:59.000Z

163

Burning actinides in very hard spectrum reactors  

SciTech Connect

The major unresolved problem in the nuclear industry is the ultimate disposition of the waste products of light water reactors. The study demonstrates the feasibility of designing a very hard spectrum actinide burner reactor (ABR). A 1100 MW/sub t/ ABR design fueled entirely with actinides reprocessed from light water reactor (LWR) wastes is proposed as both an ultimate disposal mechanism for actinides and a means of concurrently producing usable power. Actinides from discharged ABR fuel are recycled to the ABR while fission products are routed to a permanent repository. As an integral part of a large energy park, each such ABR would dispose of the waste actinides from 2 LWRs.

Robinson, A.H.; Shirley, G.W.; Prichard, A.W.; Trapp, T.J.

1978-03-20T23:59:59.000Z

164

Darlington tritium removal facility and station upgrading plant dynamic process simulation  

SciTech Connect

Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station has been operating a Tritium Removal Facility (TRF) and a D{sub 2}O station Upgrading Plant (SUP) since 1989. Both facilities were designed with a Distributed Control System (DCS) and programmable logic controllers (PLC) for process control. This control system was replaced with a DCS only, in 1998. A dynamic plant simulator was developed for the Darlington TRF (DTRF) and the SUP, as part of the computer control system replacement. The simulator was used to test the new software, required to eliminate the PLCs. The simulator is now used for operator training and testing of process control software changes prior to field installation. Dynamic simulation will be essential for the ITER isotope separation system, where the process is more dynamic than the relatively steady-state DTRF process. This paper describes the development and application of the DTRF and SUP dynamic simulator, its benefits, architecture, and the operational experience with the simulator. (authors)

Busigin, A. [NITEK USA, Inc., 6405 NW 77 PL, Parkland, FL 33067 (United States); Williams, G. I. D.; Wong, T. C. W.; Kulczynski, D.; Reid, A. [Ontario Power Generation Nuclear, Box 4000, Bowmanville, ON L1C 3Z8 (Canada)

2008-07-15T23:59:59.000Z

165

Process for the removal of acid forming gases from exhaust gases  

DOE Patents (OSTI)

Exhaust gases are treated to remove NO or NO[sub x] and SO[sub 2] by contacting the gases with an aqueous emulsion or suspension of yellow phosphorus preferably in a wet scrubber. The pressure is not critical, and ambient pressures are used. Hot water temperatures are best, but economics suggest about 50 C is attractive. The amount of yellow phosphorus used will vary with the composition of the exhaust gas, less than 3% for small concentrations of NO, and 10% or higher for concentrations above say 1000 ppm. Similarly, the pH will vary with the composition being treated, and it is adjusted with a suitable alkali. For mixtures of NO[sub x] and SO[sub 2], alkalis that are used for flue gas desulfurization are preferred. With this process, 100% of the by-products created are usable, and close to 100% of the NO or NO[sub x] and SO[sub 2] can be removed in an economic fashion. 9 figs.

Chang, S.G.; Liu, D.K.

1992-11-17T23:59:59.000Z

166

Evaluation of a Combined Cyclone & Gas Filtration System for Particulate Removal in the Gasification Process  

Science Conference Proceedings (OSTI)

The Wabash gasification facility, owned and operated by sgSolutions LLC, is one of the largest single train solid fuel gasification facilities in the world capable of transforming 2,000 tons per day of petroleum coke or 2,600 tons per day of bituminous coal into synthetic gas for electrical power generation. The Wabash plant utilizes Phillips66 proprietary E-Gas™ Gasification Process to convert solid fuels such as petroleum coke or coal into synthetic gas that is fed to a combined cycle combustion turbine power generation facility. During plant startup in 1995, reliability issues were realized in the gas filtration portion of the gasification process. To address these issues, a slipstream test unit was constructed at the Wabash facility to test various filter designs, materials and process conditions for potential reliability improvement. The char filtration slipstream unit provided a way of testing new materials, maintenance procedures, and process changes without the risk of stopping commercial production in the facility. It also greatly reduced maintenance expenditures associated with full scale testing in the commercial plant. This char filtration slipstream unit was installed with assistance from the United States Department of Energy (built under DOE Contract No. DE-FC26-97FT34158) and began initial testing in November of 1997. It has proven to be extremely beneficial in the advancement of the E-Gas™ char removal technology by accurately predicting filter behavior and potential failure mechanisms that would occur in the commercial process. After completing four (4) years of testing various filter types and configurations on numerous gasification feed stocks, a decision was made to investigate the economic and reliability effects of using a particulate removal gas cyclone upstream of the current gas filtration unit. A paper study had indicated that there was a real potential to lower both installed capital and operating costs by implementing a char cyclonefiltration hybrid unit in the E-Gas™ gasification process. These reductions would help to keep the E-Gas™ technology competitive among other coal-fired power generation technologies. The Wabash combined cyclone and gas filtration slipstream test program was developed to provide design information, equipment specification and process control parameters of a hybrid cyclone and candle filter particulate removal system in the E-Gas™ gasification process that would provide the optimum performance and reliability for future commercial use. The test program objectives were as follows: 1. Evaluate the use of various cyclone materials of construction. 2. Establish the optimal cyclone efficiency that provides stable long term gas filter operation. 3. Determine the particle size distribution of the char separated by both the cyclone and candle filters. This will provide insight into cyclone efficiency and potential future plant design. 4. Determine the optimum filter media size requirements for the cyclone-filtration hybrid unit. 5. Determine the appropriate char transfer rates for both the cyclone and filtration portions of the hybrid unit. 6. Develop operating procedures for the cyclone-filtration hybrid unit. 7. Compare the installed capital cost of a scaled-up commercial cyclone-filtration hybrid unit to the current gas filtration design without a cyclone unit, such as currently exists at the Wabash facility.

Rizzo, Jeffrey

2010-04-30T23:59:59.000Z

167

Development of an Integrated Multi-Contaminant Removal Process Applied to Warm Syngas Cleanup  

NLE Websites -- All DOE Office Websites (Extended Search)

Gasification Gasification Technologies contacts Gary J. stiegel Gasification Technology Manager National Energy Technology Laboratory 626 Cochrans Mill Road P.O. Box 10940 Pittsburgh, PA 15236 412-386-4499 gary.stiegel@netl.doe.gov Jenny tennant Project Manager National Energy Technology Laboratory 3610 Collins Ferry Road P.O. Box 880 Morgantown, WV 26507 304-285-4830 jenny.tennant@netl.doe.gov Howard Meyer Principal Project Manager Gas Technology Institute 1700 South Mount Prospect Road Des Plaines, IL 60018 847-768-0955 howard.meyer@gastechnology.org Development of an IntegrateD multI-ContamInant removal proCess applIeD to Warm syngas Cleanup Description The U.S. has more coal than any other country, and through gasification this coal can be converted into electricity, liquid fuels, chemicals or hydrogen. However,

168

Catalytic two-stage coal liquefaction process having improved nitrogen removal  

SciTech Connect

A process for catalytic multi-stage hydrogenation and liquefaction of coal to produce high yields of low-boiling hydrocarbon liquids containing low concentrations of nitogen compounds. First stage catalytic reaction conditions are 700.degree.-800.degree. F. temperature, 1500-3500 psig hydrogen partial pressure, with the space velocity maintained in a critical range of 10-40 lb coal/hr ft.sup.3 catalyst settled volume. The first stage catalyst has 0.3-1.2 cc/gm total pore volume with at least 25% of the pore volume in pores having diameters of 200-2000 Angstroms. Second stage reaction conditions are 760.degree.-870.degree. F. temperature with space velocity exceeding that in the first stage reactor, so as to achieve increased hydrogenation yield of low-boiling hydrocarbon liquid products having at least 75% removal of nitrogen compounds from the coal-derived liquid products.

Comolli, Alfred G. (Yardley, PA)

1991-01-01T23:59:59.000Z

169

Process and system for removing sulfur from sulfur-containing gaseous streams  

DOE Patents (OSTI)

A multi-stage UCSRP process and system for removal of sulfur from a gaseous stream in which the gaseous stream, which contains a first amount of H.sub.2S, is provided to a first stage UCSRP reactor vessel operating in an excess SO.sub.2 mode at a first amount of SO.sub.2, producing an effluent gas having a reduced amount of SO.sub.2, and in which the effluent gas is provided to a second stage UCSRP reactor vessel operating in an excess H.sub.2S mode, producing a product gas having an amount of H.sub.2S less than said first amount of H.sub.2S.

Basu, Arunabha (Aurora, IL); Meyer, Howard S. (Hoffman Estates, IL); Lynn, Scott (Pleasant Hill, CA); Leppin, Dennis (Chicago, IL); Wangerow, James R. (Medinah, IL)

2012-08-14T23:59:59.000Z

170

Apparatus for removing noncondensable gases from cogenerated process steam in dual fluid cheng cycle engines  

SciTech Connect

An apparatus is described for removing noncondensable gases from process steam cogenerated in a steam-injected gas turbine engine. The engine consists of: (a) a chamber; (b) compressor means for introducing air into the chamber; (c) means for introducing steam within the chamber, the steam introducing means including an automatically controlled steam injector valve and steam injection line, (d) means for heating the air and steam in the chamber, including means for combustion; (e) turbine means responsive to a mixture of air, combustion products and steam for converting the energy associated with the mixture to mechanical energy; (f) counterflow heat exchanger means, including at least superheater and evaporator sections, for transferring residual thermal energy from a mixture of air, combustion products and steam exhausted from the turbine means to incoming water and steam.

Cheng, D.Y.

1987-08-11T23:59:59.000Z

171

Flotation process for removal of precipitates from electrochemical chromate reduction unit  

DOE Patents (OSTI)

This invention is an improved form of a conventional electrochemical process for removing hexavalent chromium or other metal-ion contaminants from cooling-tower blowdown water. In the conventional process, the contaminant is reduced and precipitated at an iron anode, thus forming a mixed precipitate of iron and chromium hydroxides, while hydrogen being evolved copiously at a cathode is vented from the electrochemical cell. In the conventional process, subsequent separation of the fine precipitate has proved to be difficult and inefficient. In accordance with this invention, the electrochemical operation is conducted in a novel manner permitting a much more efficient and less expensive precipitate-recovery operation. That is, the electrochemical operation is conducted under an evolved-hydrogen partial pressure exceeding atmospheric pressure. As a result, most of the evolved hydrogen is entrained as bubbles in the blowdown in the cell. The resulting hydrogen-rich blowdown is introduced to a vented chamber, where the entrained hydrogen combines with the precipitate to form a froth which can be separated by conventional techniques. In addition to the hydrogen, two materials present in most blowdown act as flotation promoters for the precipitate. These are (1) air, with which the blowdown water becomes saturated in the course of normal cooling-tower operation, and (2) surfactants which commonly are added to cooling-tower recirculating-water systems to inhibit the growth of certain organisms or prevent the deposition of insoluble particulates.

DeMonbrun, James R. (Knoxville, TN); Schmitt, Charles R. (Oak Ridge, TN); Williams, Everett H. (Oak Ridge, TN)

1976-01-01T23:59:59.000Z

172

Process for removing halogenated aliphatic and aromatic compounds from petroleum products  

DOE Patents (OSTI)

A process for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contacting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible compound, such as, water or a polyhydroxy compound, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of water or polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the water or polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds and the low polar or nonpolar solvent are separated by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered from recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced.

Googin, John M. (Oak Ridge, TN); Napier, John M. (Oak Ridge, TN); Travaglini, Michael A. (Oliver Springs, TN)

1983-01-01T23:59:59.000Z

173

Process for removing halogenated aliphatic and aromatic compounds from petroleum products  

DOE Patents (OSTI)

A process is described for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contacting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible compound, such as, water or a polyhydroxy compound, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of water or polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the water or polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds and the low polar or nonpolar solvent are separated by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered from recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced. 1 fig.

Googin, J.M.; Napier, J.M.; Travaglini, M.A.

1983-09-20T23:59:59.000Z

174

Process for removing and detoxifying cadmium from scrap metal including mixed waste  

SciTech Connect

Cadmium-bearing scrap from nuclear applications, such as neutron shielding and reactor control and safety rods, must usually be handled as mixed waste since it is radioactive and the cadmium in it is both leachable and highly toxic. Removing the cadmium from this scrap, and converting it to a nonleachable and minimally radioactive form, would greatly simplify disposal or recycling. A process now under development will do this by shredding the scrap; leaching it with reagents which selectively dissolve out the cadmium; reprecipitating the cadmium as its highly insoluble sulfide; then fusing the sulfide into a glassy matrix to bring its leachability below EPA limits before disposal. Alternatively, the cadmium may be recovered for reuse. A particular advantage of the process is that all reagents (except the glass frit) can easily be recovered and reused in a nearly closed cycle, minimizing the risk of radioactive release. The process does not harm common metals such as aluminum, iron and stainless steel, and is also applicable to non-nuclear cadmium-bearing scrap such as nickel-cadmium batteries.

Kronberg, J.W.

1994-07-01T23:59:59.000Z

175

Actinide transmutation; fission or fusion reactors studies: a review. [In-core transmutation  

SciTech Connect

Actinide transmutation in fission or fusion reactors is considered as a possible method of radioactive waste processing. A brief summary is presented of transmutation studies completed or in progress. (DG)

Croff, A.G.

1976-01-01T23:59:59.000Z

176

Process for removal of water and silicon mu-oxides from chlorosilanes  

DOE Patents (OSTI)

A scavenger composition having utility for removal of water and silicon mu-oxide impurities from chlorosilanes, such scavenger composition comprising: (a) a support; and (b) associated with the support, one or more compound(s) selected from the group consisting of compounds of the formula: R.sub.a-x MCl.sub.x wherein: M is a metal selected from the group consisting of the monovalent metals lithium, sodium, and potassium; the divalent metals magnesium, strontium, barium, and calcium; and the trivalent metal aluminum; R is alkyl; a is a number equal to the valency of metal M; and x is a number having a value of from 0 to a, inclusive; and wherein said compound(s) of the formula R.sub.a-x MCl.sub.x have been activated for impurity-removal service by a reaction scheme selected from those of the group consisting of: (i) reaction of such compound(s) with hydrogen chloride to form a first reaction product therefrom, followed by reaction of the first reaction product with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y, wherein y is a number having a value of from 1 to 3, inclusive; and (ii) reaction of such compound(s) with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y wherein y is a number having a value of 1 to 3, inclusive. A corresponding method of making the scavenger composition, and of purifying a chlorosilane which contains oxygen and silicon mu-oxide impurities, likewise are disclosed, together with a purifier apparatus, in which a bed of the scavenger composition is disposed. The composition, purification process, and purifier apparatus of the invention have utility in purifying gaseous chlorosilanes which are employed in the semiconductor industry as silicon source reagents for forming epitaxial silicon layers.

Tom, Glenn M. (New Milford, CT); McManus, James V. (Danbury, CT)

1992-03-10T23:59:59.000Z

177

Composition, process, and apparatus, for removal of water and silicon mu-oxides from chlorosilanes  

DOE Patents (OSTI)

A scavenger composition having utility for removal of water and silicon mu-oxide impurities from chlorosilanes, such scavenger composition comprising: (a) a support; and (b) associated with the support, one or more compound(s) selected from the group consisting of compounds of the formula: R.sub.a-x MCl.sub.x wherein: M is a metal selected from the group consisting of the monovalent metals lithium, sodium, and potassium; the divalent metals magnesium, strontium, barium, and calcium; and the trivalent metal aluminum; R is alkyl; a is a number equal to the valency of metal M; and x is a number having a value from 0 to a, inclusive; and wherein said compound(s) of the formula R.sub.a-x MCl.sub.x have been activated for impurity-removal service by a reaction scheme selected from those of the group consisting of: (i) reaction of such compound(s) with hydrogen chloride to form a first reaction product therefrom, followed by reaction of the first reaction product with a chlorosilane of the formula: SiH.sub.4"y Cl.sub.y, wherein y is a number having a value of from 1 to 3, inclusive; and (ii) reaction of such compound(s) with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y wherein y is a number having a value of 1 to 3, inclusive. A corresponding method of making the scavenger composition, and of purifying a chlorosilane which contains oxygen and silicon mu-oxide impurities, likewise are disclosed, together with a purifier apparatus, in which a bed of the scavenger composition is disposed. The composition, purification process, and purifier apparatus of the invention have utility in purifying gaseous chlorosilanes which are employed in the semiconductor industry as silicon source reagents for forming epitaxial silicon layers.

Tom, Glenn M. (New Milford, CT); McManus, James V. (Danbury, CT)

1991-10-15T23:59:59.000Z

178

Process for the removal of acid forming gases from exhaust gases and production of phosphoric acid  

DOE Patents (OSTI)

Exhaust gases are treated to remove NO or NO.sub.x and SO.sub.2 by contacting the gases with an aqueous emulsion or suspension of yellow phosphorous preferably in a wet scrubber. The addition of yellow phosphorous in the system induces the production of O.sub.3 which subsequently oxidizes NO to NO.sub.2. The resulting NO.sub.2 dissolves readily and can be reduced to form ammonium ions by dissolved SO.sub.2 under appropriate conditions. In a 20 acfm system, yellow phosphorous is oxidized to yield P.sub.2 O.sub.5 which picks up water to form H.sub.3 PO.sub.4 mists and can be collected as a valuable product. The pressure is not critical, and ambient pressures are used. Hot water temperatures are best, but economics suggest about 50.degree. C. The amount of yellow phosphorus used will vary with the composition of the exhaust gas, less than 3% for small concentrations of NO, and 10% or higher for concentrations above say 1000 ppm. Similarly, the pH will vary with the composition being treated, and it is adjusted with a suitable alkali. For mixtures of NO.sub.x and SO.sub.2, alkalis that are used for flue gas desulfurization are preferred. With this process, better than 90% of SO.sub.2 and NO in simulated flue gas can be removed. Stoichiometric ratios (P/NO) ranging between 0.6 and 1.5 were obtained.

Chang, Shih-Ger (El Cerrito, CA); Liu, David K. (San Pablo, CA)

1992-01-01T23:59:59.000Z

179

Organophosphorus reagents in actinide separations: Unique tools for production, cleanup and disposal  

Science Conference Proceedings (OSTI)

Interactions of actinide ions with phosphate and organophosphorus reagents have figured prominently in nuclear science and technology, particularly in the hydrometallurgical processing of irradiated nuclear fuel. Actinide interactions with phosphorus-containing species impact all aspects from the stability of naturally occurring actinides in phosphate mineral phases through the application of the bismuth phosphate and PUREX processes for large-scale production of transuranic elements to the development of analytical separation and environment restoration processes based on new organophosphorus reagents. In this report, an overview of the unique role of organophosphorus compounds in actinide production, disposal, and environment restoration is presented. The broad utility of these reagents and their unique chemical properties is emphasized.

Nash, K. L.

2000-01-12T23:59:59.000Z

180

CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES  

SciTech Connect

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

Nash, Kenneth L.

2008-11-20T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

MODELING ION-EXCHANGE PROCESSING WITH SPHERICAL RESINS FOR CESIUM REMOVAL  

SciTech Connect

The spherical Resorcinol-Formaldehyde and hypothetical spherical SuperLig? 644 ion-exchange resins are evaluated for cesium removal from radioactive waste solutions. Modeling results show that spherical SuperLig? 644 reduces column cycling by 50% for highpotassium solutions. Spherical Resorcinol Formaldehyde performs equally well for the lowestpotassium wastes. Less cycling reduces nitric acid usage during resin elution and sodium addition during resin regeneration, therefore, significantly decreasing life-cycle operational costs. A model assessment of the mechanism behind ?cesium bleed? is also conducted. When a resin bed is eluted, a relatively small amount of cesium remains within resin particles. Cesium can bleed into otherwise decontaminated product in the next loading cycle. The bleed mechanism is shown to be fully isotherm-controlled vs. mass transfer controlled. Knowledge of residual postelution cesium level and resin isotherm can be utilized to predict rate of cesium bleed in a mostly non-loaded column. Overall, this work demonstrates the versatility of the ion-exchange modeling to study the effects of resin characteristics on processing cycles, rates, and cold chemical consumption. This evaluation justifies further development of a spherical form of the SL644 resin.

Hang, T.; Nash, C.; Aleman, S.

2012-09-19T23:59:59.000Z

182

Separations and Actinide Science -- 2005 Roadmap  

SciTech Connect

The Separations and Actinide Science Roadmap presents a vision to establish a separations and actinide science research (SASR) base composed of people, facilities, and collaborations and provides new and innovative nuclear fuel cycle solutions to nuclear technology issues that preclude nuclear proliferation. This enabling science base will play a key role in ensuring that Idaho National Laboratory (INL) achieves its long-term vision of revitalizing nuclear energy by providing needed technologies to ensure our nation's energy sustainability and security. To that end, this roadmap suggests a 10-year journey to build a strong SASR technical capability with a clear mission to support nuclear technology development. If nuclear technology is to be used to satisfy the expected growth in U.S. electrical energy demand, the once-through fuel cycle currently in use should be reconsidered. Although the once-through fuel cycle is cost-effective and uranium is inexpensive, a once-through fuel cycle requires long-term disposal to protect the environment and public from long-lived radioactive species. The lack of a current disposal option (i.e., a licensed repository) has resulted in accumulation of more than 50,000 metric tons of spent nuclear fuel. The process required to transition the current once-through fuel cycle to full-recycle will require considerable time and significant technical advancement. INL's extensive expertise in aqueous separations will be used to develop advanced separations processes. Computational chemistry will be expanded to support development of future processing options. In the intermediate stage of this transition, reprocessing options will be deployed, waste forms with higher loading densities and greater stability will be developed, and transmutation of long-lived fission products will be explored. SASR will support these activities using its actinide science and aqueous separations expertise. In the final stage, full recycle will be enabled by advanced reactors and reprocessing methods based on pyrochemical methods and by using different fuel cycles that do not readily produce plutonium. SASR will facilitate the deployment of advanced pyrochemical separation technology and support development of reprocessing of thorium-based reactor fuels.

2005-09-01T23:59:59.000Z

183

Ceramic Composition for Immobilization of Actinides  

DOE Patents (OSTI)

Disclosed is a ceramic composition for the immobilization of actinides, particularly uranium and plutonium. The ceramic is a titanate material comprising pyrochlore, brannerite and rutile.

Ebbinghaus, Bartley B.; Van Konynenburg, Richard A.; Vance, Eric R.; Stewart, Martin W.; Jostsons, Adam; Allender, Jeffrey S.; Rankin, David Thomas

1999-06-22T23:59:59.000Z

184

TAILORING INORGANIC SORBENTS FOR SRS STRONTIUM AND ACTINIDE SEPARATIONS: MODIFIED MONOSODIUM TITANATE PHASE III FINAL REPORT  

DOE Green Energy (OSTI)

This document provides a final report of Phase III testing activities for the development of modified monosodium titanate (mMST), which exhibits improved strontium and actinide removal characteristics compared to the baseline MST material. The activities included characterization of the crystalline phases present at varying temperatures, solids settling characteristics, quantification of the peroxide content; evaluation of the post-synthesis gas release under different conditions; the extent of desorption of {sup 85}Sr, Np, and Pu under washing conditions; and the effects of age and radiation on the performance of the mMST. Key findings and conclusions include the following. The peroxide content of several mMST samples was determined using iodometric titration. The peroxide content was found to decrease with age or upon extended exposure to elevated temperature. A loss of peroxide was also measured after exposure of the material to an alkaline salt solution similar in composition to the simulated waste solution. To determine if the loss of peroxide with age affects the performance of the material, Sr and actinide removal tests were conducted with samples of varying age. The oldest sample (4 years and 8 months) did show lower Sr and Pu removal performance. When compared to the youngest sample tested (1 month), the oldest sample retained only 15% of the DF for Pu. Previous testing with this sample indicated no decrease in Pu removal performance up to an age of 30 months. No loss in Np removal performance was observed for any of the aged samples, and no uptake of uranium occurred at the typical sorbent loading of 0.2 g/L. Additional testing with a uranium only simulant and higher mMST loading (3.0 g/L) indicated a 10% increase of uranium uptake for a sample aged 3 years and 8 months when compared to the results of the same sample measured at an age of 1 year and 5 months. Performance testing with both baseline-MST and mMST that had been irradiated in a gamma source to a total dose of 3.95 x 10{sup 6} R, indicated little to no affect on the performance of the material to remove Sr and actinides. Previous testing established that mMST releases oxygen gas during the synthesis, and continues to off-gas during storage post synthesis. The post-synthesis gas release rate was measured under several conditions, including varying the pH of the wash water and at elevated temperature (49 C, typical of bounding summertime storage without air conditioning). Results indicated that a high pH (basic) wash reduced the initial gas release rate, but after 2 days the release rates from all different pH washed samples were not statistically different. The gas release rate at 49 C, a temperature at which the material may be exposed to during shipping and storage, was consistently about 2.5 times higher than the rate at room temperature. All gas release results indicated that vented containers would be necessary for shipping and storage of large quantities of material. Suspension of sorbate-loaded solids into diluted solutions representing intermediate and final stages of washing for 24-hours revealed no evidence of desorption of Sr, Pu or Np from the mMST solids. Based on the results of the Phase III testing as well as that from earlier studies (Phases I and II), SRNL researchers recommend adopting the use of the mMST material for the removal of strontium and actinides from the SRS HLW supernatant liquids in the Actinide Removal Process and Salt Waste Processing Facility. Given the decrease in Sr and Pu removal performance for the mMST having an age of 4 years and 8 months, we recommend that mMST be used within 30 months of production. Furthermore we recommend that DOE provide funding to conduct pilot-scale testing of the mixing and settling characteristics of the mMST and impact, if any, on the generation of hydrogen during processing in the Defense Waste Processing Facility (DWPF).

Taylor-Pashow, K.; Hobbs, D.

2010-09-01T23:59:59.000Z

185

Demonstration of a Universal Solvent Extraction Process for the Separation of Cesium and Strontium from Actual Acidic Tank Waste at the INEEL  

Science Conference Proceedings (OSTI)

A universal solvent extraction process is being evaluated for the simultaneous separation of Cs, Sr, and the actinides from acidic high-activity tank waste at the Idaho National Engineering and Environmental Laboratory (INEEL) with the goal of minimizing the high-activity waste volume to be disposed in a deep geological repository. The universal solvent extraction process is being developed as a collaborative effort between the INEEL and the Khlopin Radium Institute in St. Petersburg, Russia. The process was recently demonstrated at the INEEL using actual radioactive, acidic tank waste in 24 stages of 2-cm-diameter centrifugal contactors located in a shielded cell facility. With the testing, removal efficiencies of 99.95%, 99.985%, and 95.2% were obtained for Cs-137, Sr-90, and total alpha, respectively. This is sufficient to reduce the activities of Cs-137 and Sr-90 to below NRC Class A LLW requirements. The total alpha removal efficiency was not sufficient to reduce the activity of the tank waste to below NRC Class A non-TRU requirements. The lower than expected removal efficiency for the actinides is due to loading of the Ph2Bu2CMPO in the universal solvent with actinides and metals (Zr, Fe, and Mo). Also, the carryover of aqueous solution (flooding) with the solvent exiting the actinide strip section and entering the wash section resulted in the recycle of the actinides back to the extraction section. This recycle of the actinides contributed to the low removal efficiency. Significant amounts of the Zr (>97.7%), Ba (>87%), Pb (>98.5%), Fe (>6.9%), Mo (19%), and K (17%) were also removed from the feed with the universal solvent extraction flowsheet.

B. N. Zaitsev (Khlopin Radium Institute); D. J. Wood (INEEL); I. V. Smirnov; J. D. Law; R. S. Herbst; T. A. Todd; V. A. Babain; V. M. Esimantovskiy; V. N. Romanovskiy

1999-08-01T23:59:59.000Z

186

EA-1404: Actinide Chemistry and Repository Science Laboratory...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

4: Actinide Chemistry and Repository Science Laboratory, Carlsbad, New Mexico EA-1404: Actinide Chemistry and Repository Science Laboratory, Carlsbad, New Mexico SUMMARY This EA...

187

Process and apparatus for adding and removing particles from pressurized reactors  

DOE Patents (OSTI)

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Milligan, John D. (Little Silver, NJ)

1983-01-01T23:59:59.000Z

188

Processing & Applications IV: Radioactive Metals Processing  

Science Conference Proceedings (OSTI)

The compatibility between erbium oxide and molten cerium, as a surrogate for ... A key process step involving molten salt electrorefining to separate actinides ...

189

Actinide partitioning-transmutation program final report. I. Overall assessment  

SciTech Connect

This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of /sup 99/Tc and /sup 129/I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted.

Croff, A.G.; Blomeke, J.O.; Finney, B.C.

1980-06-01T23:59:59.000Z

190

Method of removing the effects of electrical shorts and shunts created during the fabrication process of a solar cell  

DOE Patents (OSTI)

A method of removing the effects of electrical shorts and shunts created during the fabrication process and improving the performance of a solar cell with a thick film cermet electrode opposite to the incident surface by applying a reverse bias voltage of sufficient magnitude to burn out the electrical shorts and shunts but less than the break down voltage of the solar cell.

Nostrand, Gerald E. (Jamesburg, NJ); Hanak, Joseph J. (Lawrenceville, NJ)

1979-01-01T23:59:59.000Z

191

The removal of mercury from solid mixed waste using chemical leaching processes  

Science Conference Proceedings (OSTI)

The focus of this research was to evaluate chemical leaching as a technique to treat soils, sediments, and glass contaminated with either elemental mercury or a combination of several mercury species. Potassium iodide/iodine solutions were investigated as chemical leaching agents for contaminated soils and sediments. Clean, synthetic soil material and surrogate storm sewer sediments contaminated with mercury were treated with KI/I{sub 2} solutions. It was observed that these leaching solutions could reduce the mercury concentration in soil and sediments by 99.8%. Evaluation of selected posttreatment sediment samples revealed that leachable mercury levels in the treated solids exceeded RCRA requirements. The results of these studies suggest that KI/I{sub 2} leaching is a treatment process that can be used to remove large quantities of mercury from contaminated soils and sediments and may be the only treatment required if treatment goals are established on Hg residual concentrations in solid matrices. Fluorescent bulbs were used to simulate mercury contaminated glass mixed waste. To achieve mercury contamination levels similar to those found in larger bulbs such as those used in DOE facilities a small amount of Hg was added to the crushed bulbs. The most effective agents for leaching mercury from the crushed fluorescent bulbs were KI/I{sub 2}, NaOCl, and NaBr + acid. Radionuclide surrogates were added to both the EPA synthetic soil material and the crushed fluorescent bulbs to determine the fate of radionuclides following chemical leaching with the leaching agents determined to be the most promising. These experiments revealed that although over 98% of the dosed mercury solubilized and was found in the leaching solution, no Cerium was measured in the posttreatment leaching solution. This finding suggest that Uranium, for which Ce was used as a surrogate, would not solubilize during leaching of mercury contaminated soil or glass.

Gates, D.D.; Chao, K.K.; Cameron, P.A.

1995-07-01T23:59:59.000Z

192

Processes for Removal and Immobilization of 14C, 129I, and 85Kr  

SciTech Connect

This is a white paper covering the results of a literature search and preliminary experiments on materials and methods to remove and immobilize gaseous radionuclided that come from the reprocessing of spent nuclear fuel.

Strachan, Denis M.; Bryan, Samuel A.; Henager, Charles H.; Levitskaia, Tatiana G.; Matyas, Josef; Thallapally, Praveen K.; Scheele, Randall D.; Weber, William J.; Zheng, Feng

2009-10-05T23:59:59.000Z

193

CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE D&D OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT (PFP)  

SciTech Connect

This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place.

HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

2007-01-25T23:59:59.000Z

194

Demonstration of an optimized TRUEX flowsheet for partitioning of actinides from actual ICPP sodium-bearing waste using centrifugal contactors in a shielded cell facility  

Science Conference Proceedings (OSTI)

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of the actinides from acidic radioactive wastes stored at the ICPP. These efforts have culminated in recent demonstrations of the TRUEX process with actual tank waste. The first demonstration was performed in 1996 using 24 stages of 2-cm diameter centrifugal contactors and waste from tank WM-183. Based on the results of this flowsheet demonstration, the flowsheet was optimized and a second flowsheet demonstration was performed. This test also was performed using 2-cm diameter centrifugal contactors and waste from tank WM-183. However, the total number of contactor stages was reduced from 24 to 20. Also, the concentration of HEDPA in the strip solution was reduced from 0.04 M to 0.01 M in order to minimize the amount of phosphate in the HLW fraction, which would be immobilized into a glass waste form. This flowsheet demonstration was performed using centrifugal contactors installed in the shielded hot cell at the ICPP Remote Analytical Laboratory. The flowsheet tested consisted of six extraction stages, four scrub stages, six strip stages, two solvent was stages, and two acid rinse stages. An overall removal efficiency of 99.79% was obtained for the actinides. As a result, the activity of the actinides was reduced from 540 nCi/g in the feed to 0.90 nCi/g in the aqueous raffinate, which is well below the NRC Class A LLW requirement of 10 nCi/g for non-TRU waste. Removal efficiencies of 99.84%, 99.97%, 99.97%, 99.85%, and 99.76% were obtained for {sup 241}Am, {sup 238}Pu, {sup 239}Pu, {sup 235}U, and {sup 238}U, respectively.

Law, J.D.; Brewer, K.N.; Herbst, R.S.; Todd, T.A.; Olson, L.G.

1998-01-01T23:59:59.000Z

195

Separation of actinides from lanthanides  

DOE Patents (OSTI)

An organic extracting solution and an extraction method useful for separating elements of the actinide series of the periodic table from elements of the lanthanide series, where both are in trivalent form is described. The extracting solution consists of a primary ligand and a secondary ligand, preferably in an organic solvent. The primary ligand is a substituted monothio-1,3-dicarbonyl, which includes a substituted 4-acyl-2-pyrazolin-5-thione, such as 4-benzoyl-2,4- dihydro-5-methyl-2-phenyl-3H-pyrazol-3-thione (BMPPT). The secondary ligand is a substituted phosphine oxide, such as trioctylphosphine oxide (TOPO).

Smith, B.F.; Jarvinen, G.D.; Ryan, R.R.

1988-03-31T23:59:59.000Z

196

Removal of H{sub2}S from geothermal steam by catalytic oxidation process: bench scale testing results. Interim report  

SciTech Connect

A process was investigated to remove hydrogen sulfide (H{sub2}S) from geothermal steam. This process is an upstream steam treatment process which utilizes a catalytic oxidation reaction to convert H{sub2}S in geothermal steam to water vapor and sulfur. The process consists of passing geothermal steam, containing H{sub2}S and other noncondensible gases, through fixed beds of activated carbon catalyst. Oxygen is provided by injection of air or oxygen upstream of the catalyst beds. The treated steam, with H{sub2}S being almost completely removed, passes to steam turbines for power generation. The elemental sulfur produced deposits on the catalyst surface and is retained. The catalyst activity decreases gradually with sulfur accumulation. Sulfur removal, and catalyst regeneration, is accomplished by solvent extraction. Sulfur is recovered from solvent by evaporation/crystallization. Bench scale experimental work on this process was performed to determine its performance and limits of applicability to power generation systems employing geothermal steam. The bench scale system employed a one-inch diameter reactor, a steam supply with controlled temperature and pressure, an injection system for adding {Hsub2}S and other gases at controlled rates, and instrumentation for control and measurement of temperatures, pressures, flow rates and presssure drop. H{sub2}S and other analyses were performed by wet chemistry techniques.

Li, C.T.; Brouns, R.A.

1978-11-01T23:59:59.000Z

197

The use of mathematical modeling and pilot plant testing to develop a new biological phosphorus and nitrogen removal process  

Science Conference Proceedings (OSTI)

A mechanistic mathematical model for carbon oxidation, nitrogen removal, and enhanced biological phosphorus removal was used to develop the Step Bio-P process, a new biological phosphorus and nitrogen removal process with a step-feed configuration. A 9,000-L pilot plant with diurnally varying influent process loading rates was operated to verify the model results and to optimize the Step Bio-P process for application at the lethbridge, Alberta, Canada, wastewater treatment plant. The pilot plant was operated for 10 months. An automatic on-line data acquisition system with multiple sampling and metering points for dissolved oxygen, mixed liquor suspended solids, ammonia-nitrogen, nitrate-nitrogen, ortho-phosphate, and flow rates was used. A sampling program to obtain off-line data was carried out to verify the information from the on-line system and monitor additional parameters. The on-line and off-line data were used to recalibrate the model, which was used as an experimental design and process optimization tool.

Nolasco, D.A.; Daigger, G.T.; Stafford, D.R.; Kaupp, D.M.; Stephenson, J.P.

1998-09-01T23:59:59.000Z

198

Application of extraction chromatography to actinide decontamination of hydrochloric acid effluent streams  

SciTech Connect

Extraction chromatography is under development as a method to lower actinide activity levels in effluent steams. Successful application of this technique for radioactive liquid waste treatment would provide a low activity feed stream for HCl recycle, reduce the loss of radioactivity to the environment in aqueous effluents, and would lower the quantity and reduce the hazard of the associated solid waste. The extraction of Pu and Am from HCl solutions was examined for several commercial and laboratory-produced sorbed resin materials. Inert supports included silica and polymer beads of differing mesh sizes. The support material was coated with either n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (O-CMPO) or di-(4-t-butylphenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (D-CMPO) as an extractant, and using either tributyl phosphate (TBP) or diamyl amylphosphonate (DAAP) as a diluent. Solutions tested were effluent streams generated by ion exchange and solvent extraction recovery of Pu. A finer mesh silica support material demonstrated advantages in removal of trivalent Am in some tests, but also showed a tendency toward plugging and channeling as column sizes and flow rates were increased. Larger bead sizes showed better physical properties as the process was scaled up to removal of gram quantities of Am from large effluent volumes. The ratio of extractant to diluent also appeared to play a role in the retention of Am. In direct comparative studies, when loaded on identical supports and diluent conditions, D-CMPO demonstrated better Am retention than O-CMPO from HCl process effluents.

Schulte, L.D.; McKee, S.D.; Salazar, R.R.

1996-05-01T23:59:59.000Z

199

Fate of soluble uranium in the I{sub 2}/KI leaching process for mercury removal  

SciTech Connect

General Electric Corporation has developed an extraction and recovery system for mercury, based upon the use of iodine (oxidant) and iodide ion (complexing agent). This system has been proposed for application to select mercury-contaminated mixed waste (i.e., waste containing radionuclides as well as other hazardous constituents), which have been generated by historic activities in support of US Department of Energy (DOE) missions. This system is compared to a system utilizing hypochlorite and chloride ions for removal of mercury and uranium from a sample of authentic mixed waste sludge. Relative to the hypochlorite (bleach) system, the iodine system mobilized more mercury and less uranium from the sludge. An engineering flowsheet has been developed to treat spent iodine-containing extraction medium, allowing the system to be recycled. The fate of soluble uranium in this series of treatment unit operations was monitored by tracing isotopically-enriched uranyl ion into simulated spent extraction medium. Treatment with use of elemental iron is shown to remove > 85% of the traced uranium while concurrently reducing excess iodine to the iodide ion. The next unit operation, adjustment of the solution pH to a value near 12 by the addition of lime slurry to form a metal-laden sludge phase (an operation referred to as lime-softening), removed an additional 57% of soluble uranium activity, for an over-all removal efficiency of {approximately} 96%. However, the precipitated solids did not settle well, and some iodide reagent is held up in the wet filtercake.

Bostick, W.D.; Davis, W.H.; Jarabek, R.J. [East Tennessee Technology Park, Oak Ridge, TN (United States). Materials and Chemistry Lab.

1997-09-01T23:59:59.000Z

200

Precipitation process for the removal of technetium values from nuclear waste solutions  

DOE Patents (OSTI)

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

Walker, D.D.; Ebra, M.A.

1985-11-21T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Summary - Salt Waste Processing Facility Design at the Savannah River Site  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Salt Waste Processing Facility Salt Waste Processing Facility ETR Report Date: November 2006 ETR-4 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Salt Waste Processing Facility Design at the Savannah River Site (SRS) Why DOE-EM Did This Review The Salt Waste Processing Facility (SWPF) is intended to remove and concentrate the radioactive strontium (Sr), actinides, and cesium (Cs) from the bulk salt waste solutions in the SRS high-level waste tanks. The sludge and strip effluent from the SWPF that contain concentrated Sr, actinide, and Cs wastes will be sent to the SRS Defense Waste Processing Facility (DWPF), where they will be vitrified. The decontaminated salt solution (DSS) that is left after removal of the highly

202

Fusion Techniques for the Oxidation of Refractory Actinide Oxides  

Science Conference Proceedings (OSTI)

Small-scale experiments were performed to demonstrate the feasibility of fusing refractory actinide oxides with a series of materials commonly used to decompose minerals, glasses, and other refractories as a pretreatment to dissolution and subsequent recovery operations. In these experiments, 1-2 g of plutonium or neptunium oxide (PuO2 or NpO2) were calcined at 900 degrees Celsius, mixed and heated with the fusing reagent(s), and dissolved. For refractory PuO2, the most effective material tested was a lithium carbonate (Li2CO3)/sodium tetraborate (Na2B4O7) mixture which aided in the recovery of 90 percent of the plutonium. The fused product was identified as a lithium plutonate (Li3PuO4) by x-ray diffraction. The use of a Li2CO3/Na2B4O7 mixture to solubilize high-fired NpO2 was not as effective as demonstrated for refractory PuO2. In a small-scale experiment, 25 percent of the NpO2 was oxidized to a neptunium (VI) species that dissolved in nitric acid. The remaining neptunium was then easily recovered from the residue by fusing with sodium peroxide (Na2O2). Approximately 70 percent of the neptunium dissolved in water to yield a basic solution of neptunium (VII). The remainder was recovered as a neptunium (VI) solution by dissolving the residue in 8M nitric acid. In subsequent experiments with Na2O2, the ratio of neptunium (VII) to (VI) was shown to be a function of the fusion temperature, with higher temperatures (greater than approximately 400 degrees C) favoring the formation of neptunium (VII). The fusion of an actual plutonium-containing residue with Na2O2 and subsequent dissolution was performed to demonstrate the feasibility of a pretreatment process on a larger scale. Sodium peroxide was chosen due to the potential of achieving higher actinide recoveries from refractory materials. In this experiment, nominally 10 g of a graphite-containing residue generated during plutonium casting operations was initially calcined to remove the graphite. Removal of combustible material prior to a large-scale fusion with Na2O2 is needed due to the large amount of heat liberated during oxidation. Two successive fusions using the residue from the calcination and the residue generated from the initial dissolution allowed recovery of 98 percent of the plutonium. The fusion of the residue following the first dissolution was performed at a higher temperature (600 degrees Celsius versus 450 degrees Celsius during the first fusion). The ability to recover most of the remaining plutonium from the residue suggest the oxidation efficiency of the Na2O2 fusion improves with higher temperatures similar to results observed with NpO2 fusion.

Rudisill, T.S.

1999-04-15T23:59:59.000Z

203

Complexation of lanthanides and actinides by acetohydroxamic acid  

Science Conference Proceedings (OSTI)

Acetohydroxamic acid (AHA) has been proposed as a suitable reagent for the complexant-based, as opposed to reductive, stripping of plutonium and neptunium ions from the tributylphosphate solvent phase in advanced PUREX or UREX processes designed for future nuclear-fuel reprocessing. Stripping is achieved by the formation of strong hydrophilic complexes with the tetravalent actinides in nitric acid solutions. To underpin such applications, knowledge of the complexation constants of AHA with all relevant actinide (5f) and lanthanide (4f) ions is therefore important. This paper reports the determination of stability constants of AHA with the heavier lanthanide ions (Dy-Yb) and also U(IV) and Th(IV) ions. Comparisons with our previously published AHA stability-constant data for 4f and 5f ions are made. (authors)

Taylor, R.J. [British Technology Centre, Nexia Solutions, Sellafield, Seascale, CA20 1PG (United Kingdom); Sinkov, S.I. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Choppin, G.R. [Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL (United States)

2008-07-01T23:59:59.000Z

204

Development of an Integrated Multicontaminant Removal Process Applied to Warm Syngas Cleanup for Coal-Based Advanced Gasification Systems  

NLE Websites -- All DOE Office Websites (Extended Search)

an Integrated an Integrated Multicontaminant Removal Process Applied to Warm Syngas Cleanup for Coal-Based Advanced Gasification Systems Background The U.S. has more coal than any other country, and it can be converted through gasification into electricity, liquid fuels, chemicals, or hydrogen. However, for coal gasification to become sufficiently competitive to benefit the U.S. economy and help reduce our dependence on foreign fuels, gasification costs must be reduced

205

Process for the conversion of and aqueous biomass hydrolyzate into fuels or chemicals by the selective removal of fermentation inhibitors  

DOE Patents (OSTI)

A process of making a fuel or chemical from a biomass hydrolyzate is provided which comprises the steps of providing a biomass hydrolyzate, adjusting the pH of the hydrolyzate, contacting a metal oxide having an affinity for guaiacyl or syringyl functional groups, or both and the hydrolyzate for a time sufficient to form an adsorption complex; removing the complex wherein a sugar fraction is provided, and converting the sugar fraction to fuels or chemicals using a microorganism.

Hames, Bonnie R. (Westminster, CO); Sluiter, Amie D. (Arvada, CO); Hayward, Tammy K. (Broomfield, CO); Nagle, Nicholas J. (Broomfield, CO)

2004-05-18T23:59:59.000Z

206

THE RELATIONSHIP BETWEEN THE RADIATION SURVEY AND SITE INVESTIGATION PROCESS, THE CERCLA REMEDIAL OR REMOVAL  

E-Print Network (OSTI)

RCRA 10/89 closure approved 9/95 9204-3 0928-U 1966 1989 200 Gasoline Removed RIR, closure NA NA 5 oil Closed in RCRA Closure letter Site place subtitle C 1994 monitored 7560 40 Unknown Unknown 1000 9722-6 2312-U 1987 1994 550 Diesel Inert filled CR (4/95) NA Closure approval 2/95 (6/96) 9722-5 2313-U

207

Behavior of actinides in the Integral Fast Reactor fuel cycle  

SciTech Connect

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

1994-06-01T23:59:59.000Z

208

SPECIFIC SEQUESTERING AGENTS FOR THE ACTINIDES  

E-Print Network (OSTI)

for the actinides. Two tetra-catechol chelating agents wereprotons. The acidity of the catechol groups can be increased43 and 46. Figure 4. catechol aqueous solution, on a hanging

Raymond, Kenneth N.

2013-01-01T23:59:59.000Z

209

BWR Assembly Optimization for Minor Actinide Recycling  

Science Conference Proceedings (OSTI)

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22T23:59:59.000Z

210

Actinide minimization using pressurized water reactors  

E-Print Network (OSTI)

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01T23:59:59.000Z

211

Electronic structure and correlation effects in actinides  

Science Conference Proceedings (OSTI)

This report consists of the vugraphs given at a conference on electronic structure. Topics discussed are electronic structure, f-bonding, crystal structure, and crystal structure stability of the actinides and how they are inter-related.

Albers, R.C.

1998-12-01T23:59:59.000Z

212

Supercritical Carbon Dioxide-Soluble Ligands for Extracting Actinide Metal Ions from Porous Solids  

SciTech Connect

Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effective method for the removal of radioactive contaminants would release this valuable material for salvage. The objective of this project is to develop novel, substituted diphosphonic acid ligands that can be used for supercritical carbon dioxide extraction of actinide ions from solid wastes. Specifically, selected diphosphonic acids, which are known to form extremely stable complexes with actinides in aqueous and organic solution, are to be rendered carbon dioxide-soluble by the introduction of appropriate alkyl- or silicon-containing substituents. The metal complexation chemistry of these new ligands in SC-CO2 will then be investigated and techniques for their use in actinide extraction from porous solids developed.

Dietz, Mark L.

2001-06-01T23:59:59.000Z

213

Characterization of Actinides in Simulated Alkaline Tank Waste Sludges and Leachates  

SciTech Connect

Removal of waste-limiting components of sludge (Al, Cr, S, P) in underground tanks at Hanford by treatment with concentrated alkali has proven less efficacious for Al and Cr removal than had been hoped. More aggressive treatments of sludges, for example, contact with oxidants targeting Cr(III), have been tested in a limited number of samples and found to improve leaching efficiency for Cr. Oxidative alkaline leaching can be expected to have at best a secondary influence on the mobilization of Al. Our earlier explorations of Al leaching from sludge simulants indicated acidic and complexometric leaching can improve Al dissolution. Unfortunately, treatments of sludge samples with oxidative alkaline, acidic or complexing leachates produce conditions under which normally insoluble actinide ions (e.g., Am3+, Pu4+, Np4+) can be mobilized to the solution phase. Few experimental or meaningful theoretical studies of actinide chemistry in strongly alkaline, strongly oxidizing solutions have been completed. Unfortunately, extrapolation of the more abundant acid phase thermodynamic data to these radically different conditions provides limited reliable guidance for predicting actinide speciation in highly salted alkaline solutions. In this project, we are investigating the fundamental chemistry of actinides and important sludge components in sludge simulants and supernatants under representative oxidative leaching conditions. We are examining the potential impact of acidic or complexometric leaching with concurrent secondary separations on Al removal from sludges. Finally, a portion of our research is directed at the control of polyvalent anions (SO4=, CrO4=, PO43-) in waste streams destined for vitrification. Our primary objective is to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop. We expect to identify those components of sludges that are likely to be problematic in the application of oxidative, acidic, and complexometric leaching protocols.

Nash, Kenneth L.

2005-06-01T23:59:59.000Z

214

Lattice effects in the light actinides  

Science Conference Proceedings (OSTI)

The light actinides show a variety of lattice effects that do not normally appear in other regions of the periodic table. The article will cover the crystal structures of the light actinides, their atomic volumes, their thermal expansion behavior, and their elastic behavior as reflected in recent thermal vibration measurements made by neutron diffraction. A discussion of the melting points will be given in terms of the thermal vibration measurements. Pressure effects will be only briefly indicated.

Lawson, A.C.; Cort, B.; Roberts, J.A.; Bennett, B.I.; Brun, T.O.; Dreele, R.B. von [Los Alamos National Lab., NM (United States); Richardson, J.W. Jr. [Argonne National Lab., IL (United States)

1998-12-31T23:59:59.000Z

215

Russian refiner tests new one-stage H[sub 2]S removal process  

Science Conference Proceedings (OSTI)

The Institute of Catalysis, Novosibirsk, Russia, has developed a new technology for purifying gas streams containing hydrogen sulfide. The one-stage process was tested at BashSKTP Concern Grozneftekhim's refinery in Ufa, Russia, near the southern Ural Mountains. In a pilot-size reactor, the process achieved 99% conversion of total H[sub 2]S and 98% selectivity to sulfur. The process and test results are described briefly.

Not Available

1994-03-07T23:59:59.000Z

216

Demonstration of a Universal Solvent Extraction Process for the Separation of Cesium and Strontium from Actual Acidic Tank Waste at the INEEL  

Science Conference Proceedings (OSTI)

A universal solvent extraction process is being evaluated for the simultaneous separation of Cs, Sr, and the actinides from acidic high-activity tank waste at the Idaho National Engineering and Environmental Laboratory (INEEL) with the goal of minimizing the high-activity waste volume to be disposed in a deep geological repository. The universal solvent extraction process is being developed as a collaborative effort between the INEEL and the Khlopin Radium Institute in St. Petersburg, Russia. The process was recently demonstrated at the INEEL using actual radioactive, acidic tank waste in 24 stages of 2-cm diameter centrifugal contactors located in a shielded cell facility. With this testing, removal efficiencies of 99.95%, 99.985%, and 95.2% were obtained for 137 Cs, 90 Sr, and total alpha, respectively. This is sufficient to reduce the activities of 137 Cs and 90 Sr to below NRC Class A LLW requirements. The total alpha removal efficiency was not sufficient to reduce the activity of the tank waste to below NRC Class A non-TRU requirements. The lower than expected removal efficiency for the actinides is due to loading of the Ph2Bu2CMPO in the universal solvent exiting the actinide strip section and entering the wash section resulted in the recycle of the actinides back to the extraction section. This recycle of the actinides contributed to the low removal efficiency. Significant amounts of the Zr (>97.7%), Ba (>87%), Pb (>98.5%), Fe (6.9%), Mo (19%), and K (17%) were also removed from the feed with the universal solvent extraction flowsheet.

Law, Jack Douglas; Herbst, Ronald Scott; Todd, Terry Allen; Brewer, Ken Neal; Romanovskiy, V.N.; Esimantovskiy, V.M.; Smirnov, I.V.; Babain, V.A.; Zaitsev, B.N.

1999-09-01T23:59:59.000Z

217

MINOR ACTINIDE SEPARATIONS USING ION EXCHANGERS OR IONIC LIQUIDS  

SciTech Connect

This project seeks to determine if (1) inorganic-based ion exchange materials or (2) electrochemical methods in ionic liquids can be exploited to provide effective Am and Cm separations. Specifically, we seek to understand the fundamental structural and chemical factors responsible for the selectivity of inorganic-based ion-exchange materials for actinide and lanthanide ions. Furthermore, we seek to determine whether ionic liquids can serve as the electrolyte that would enable formation of higher oxidation states of Am and other actinides. Experiments indicated that pH, presence of complexants and Am oxidation state exhibit significant influence on the uptake of actinides and lanthanides by layered sodium titanate and hybrid zirconium and tin phosphonate ion exchangers. The affinity of the ion exchangers increased with increasing pH. Greater selectivity among Ln(III) ions with sodium titanate materials occurs at a pH close to the isoelectric potential of the ion exchanger. The addition of DTPA decreased uptake of Am and Ln, whereas the addition of TPEN generally increases uptake of Am and Ln ions by sodium titanate. Testing confirmed two different methods for producing Am(IV) by oxidation of Am(III) in ionic liquids (ILs). Experimental results suggest that the unique coordination environment of ionic liquids inhibits the direct electrochemical oxidation of Am(III). The non-coordinating environment increases the oxidation potential to a higher value, while making it difficult to remove the inner coordination of water. Both confirmed cases of Am(IV) were from the in-situ formation of strong chemical oxidizers.

Hobbs, D.; Visser, A.; Bridges, N.

2011-09-20T23:59:59.000Z

218

Method for extracting lanthanides and actinides from acid solutions by modification of Purex solvent  

DOE Patents (OSTI)

A process has been developed for the extraction of multivalent lanthanide and actinide values from acidic waste solutions, and for the separation of these values from fission product and other values, which utilizes a new series of neutral bi-functional extractants, the alkyl(phenyl)-N, N-dialkylcarbamoylmethylphosphine oxides, in combination with a phase modifier to form an extraction solution. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions.

Horwitz, E.P.; Kalina, D.G.

1984-05-21T23:59:59.000Z

219

Low Quality Natural Gas Sulfur Removal and Recovery CNG Claus Sulfur Recovery Process  

Science Conference Proceedings (OSTI)

Increased use of natural gas (methane) in the domestic energy market will force the development of large non-producing gas reserves now considered to be low quality. Large reserves of low quality natural gas (LQNG) contaminated with hydrogen sulfide (H{sub 2}S), carbon dioxide (CO{sub 2}) and nitrogen (N) are available but not suitable for treatment using current conventional gas treating methods due to economic and environmental constraints. A group of three technologies have been integrated to allow for processing of these LQNG reserves; the Controlled Freeze Zone (CFZ) process for hydrocarbon / acid gas separation; the Triple Point Crystallizer (TPC) process for H{sub 2}S / C0{sub 2} separation and the CNG Claus process for recovery of elemental sulfur from H{sub 2}S. The combined CFZ/TPC/CNG Claus group of processes is one program aimed at developing an alternative gas treating technology which is both economically and environmentally suitable for developing these low quality natural gas reserves. The CFZ/TPC/CNG Claus process is capable of treating low quality natural gas containing >10% C0{sub 2} and measurable levels of H{sub 2}S and N{sub 2} to pipeline specifications. The integrated CFZ / CNG Claus Process or the stand-alone CNG Claus Process has a number of attractive features for treating LQNG. The processes are capable of treating raw gas with a variety of trace contaminant components. The processes can also accommodate large changes in raw gas composition and flow rates. The combined processes are capable of achieving virtually undetectable levels of H{sub 2}S and significantly less than 2% CO in the product methane. The separation processes operate at pressure and deliver a high pressure (ca. 100 psia) acid gas (H{sub 2}S) stream for processing in the CNG Claus unit. This allows for substantial reductions in plant vessel size as compared to conventional Claus / Tail gas treating technologies. A close integration of the components of the CNG Claus process also allow for use of the methane/H{sub 2}S separation unit as a Claus tail gas treating unit by recycling the CNG Claus tail gas stream. This allows for virtually 100 percent sulfur recovery efficiency (virtually zero SO{sub 2} emissions) by recycling the sulfur laden tail gas to extinction. The use of the tail gas recycle scheme also deemphasizes the conventional requirement in Claus units to have high unit conversion efficiency and thereby make the operation much less affected by process upsets and feed gas composition changes. The development of these technologies has been ongoing for many years and both the CFZ and the TPC processes have been demonstrated at large pilot plant scales. On the other hand, prior to this project, the CNG Claus process had not been proven at any scale. Therefore, the primary objective of this portion of the program was to design, build and operate a pilot scale CNG Claus unit and demonstrate the required fundamental reaction chemistry and also demonstrate the viability of a reasonably sized working unit.

Klint, V.W.; Dale, P.R.; Stephenson, C.

1997-10-01T23:59:59.000Z

220

Salt Waste Processing Initiatives  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Patricia Suggs Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives 2 Overview * Current SRS Liquid Waste System status * Opportunity to accelerate salt processing - transformational technologies - Rotary Microfiltration (RMF) and Small Column Ion Exchange (SCIX) - Actinide Removal Process/Modular Caustic Side Solvent Extraction (ARP/MCU) extension with next generation extractant - Salt Waste Processing Facility (SWPF) performance enhancement - Saltstone enhancements * Life-cycle impacts and benefits 3 SRS Liquid Waste Total Volume >37 Million Gallons (Mgal) Total Curies 183 MCi (51% ) 175 MCi (49% ) >358 Million Curies (MCi) Sludge 34.3 Mgal (92% ) 3.0 Mgal (8%)

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Regenerable process for the selective removal of sulfur dioxide from effluent gases  

SciTech Connect

A regenerable process is claimed for scrubbing SO/sub 2/ from effluent gases using an aqueous alkanolamine and the corresponding sulfite as the solvent, such amine having a boiling point below about 250/sup 0/ C. At one atmosphere pressure and wherein the alkanolamine solutions containing heat stable salts (Hss) is regenerated by alkali addition, crystallization and vacuum distillation of the amine.

Atwood, G.R.; Kosseim, A.J.; Sokolik, J.E.

1983-06-21T23:59:59.000Z

222

Microbial removal of nitrogen oxides from flue gas: The BioDeNOx-process  

E-Print Network (OSTI)

W) facilities. NOx levels below 60 ppm (7% O2) have been reliably achieved, which is a reduction of 70% below combustion controls to maximize NOx reduction and minimize ammonia slip. A simplified version of the process forward in the reduction of NOx emissions from EfW facilities. INTRODUCTION Emissions from U.S. Energy

Dekker, Cees

223

Nonaqueous method for dissolving lanthanide and actinide metals  

DOE Patents (OSTI)

Lanthanide and actinide beta-diketonate complex molecular compounds are produced by reacting a beta-diketone compound with a lanthanide or actinide element in the elemental metallic state in a mixture of carbon tetrachloride and methanol.

Crisler, L.R.

1975-11-11T23:59:59.000Z

224

The carbon footprint analysis of wastewater treatment plants and nitrous oxide emissions from full-scale biological nitrogen removal processes in Spain  

E-Print Network (OSTI)

This thesis presents a general model for the carbon footprint analysis of advanced wastewater treatment plants (WWTPs) with biological nitrogen removal processes, using a life cycle assessment (LCA) approach. Literature ...

Xu, Xin, S.M. Massachusetts Institute of Technology

2013-01-01T23:59:59.000Z

225

Process for removal of polynuclear aromatics from a hydrocarbon in an endothermic reformer reaction system  

Science Conference Proceedings (OSTI)

A process is described for reforming a hydrocarbon in a multi-stage endothermic reforming series of catalytic reforming reactors where the hydrocarbon is passed through the series of catalytic reforming reactors to form a reformate. The hydrocarbon is heated prior to entry to the next catalytic reforming reactor in the series, which process comprises contact of the hydrocarbon intermediate from the series of catalytic reforming reactors containing reforming catalyst with a polynuclear aromatic adsorbent to adsorb at least a portion of the polynuclear aromatic content from the hydrocarbon prior to entry to each of the next catalytic reforming reactor in the series and recovering a reformate from the last catalytic reforming reactor in the series, the recovered reformate having a reduced content of polynuclear aromatics.

Ngan, D.Y.

1989-02-14T23:59:59.000Z

226

Application of chemical structure and bonding of actinide oxide materials for forensic science  

SciTech Connect

We are interested in applying our understanding of actinide chemical structure and bonding to broaden the suite of analytical tools available for nuclear forensic analyses. Uranium- and plutonium-oxide systems form under a variety of conditions, and these chemical species exhibit some of the most complex behavior of metal oxide systems known. No less intriguing is the ability of AnO{sub 2} (An: U, Pu) to form non-stoichiometric species described as AnO{sub 2+x}. Environmental studies have shown the value of utilizing the chemical signatures of these actinide oxide materials to understand transport following release into the environment. Chemical speciation of actinide-oxide samples may also provide clues as to the age, source, or process history of the material. The scientific challenge is to identify, measure and understand those aspects of speciation of actinide analytes that carry information about material origin and history most relevant to forensics. Here, we will describe our efforts in material synthesis and analytical methods development that we will use to provide the fundamental science to characterize actinide oxide molecular structures for forensic science. Structural properties and initial results to measure structural variability of uranium oxide samples using synchrotron-based X-ray Absorption Fine Structure will be discussed.

Wilkerson, Marianne Perry [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

227

FIELD DEMONSTRATION OF A MEMBRANE PROCESS TO RECOVER HEAVY HYDROCARBONS AND TO REMOVE WATER FROM NATURAL GAS  

SciTech Connect

The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world conditions would convince industry users of the efficiency and reliability of the process. The system has been designed and fabricated by Membrane Technology and Research, Inc. (MTR) and will be installed and operated at British Petroleum (BP)-Amoco's Pascagoula, MS plant. The Gas Research Institute will partially support the field demonstration and BP-Amoco will help install the unit and provide onsite operators and utilities. The gas processed by the membrane system will meet pipeline specifications for dewpoint and Btu value and can be delivered without further treatment to the pipeline. Based on data from prior membrane module tests, the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. At the end of this demonstration project the process will be ready for commercialization. The route to commercialization will be developed during this project and may involve collaboration with other companies already servicing the natural gas processing industry.

R. Baker; R. Hofmann; K.A. Lokhandwala

2003-02-14T23:59:59.000Z

228

Field Demonstration of a Membrane Process to Recover Heavy Hydrocarbons and to Remove Water from Natural Gas  

SciTech Connect

The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world conditions would convince industry users of the efficiency and reliability of the process. The system has been designed and fabricated by Membrane Technology and Research, Inc. (MTR) and will be installed and operated at British Petroleum (BP)-Amoco's Pascagoula, MS plant. The Gas Research Institute will partially support the field demonstration and BP-Amoco will help install the unit and provide onsite operators and utilities. The gas processed by the membrane system will meet pipeline specifications for dewpoint and BTU value and can be delivered without further treatment to the pipeline. Based on data from prior membrane module tests, the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. At the end of this demonstration project the process will be ready for commercialization. The route to commercialization will be developed during this project and may involve collaboration with other companies already servicing the natural gas processing industry.

R. Baker; T. Hofmann; K. A. Lokhandwala

2004-09-29T23:59:59.000Z

229

Field Demonstration of a Membrane Process to Recover Heavy Hydrocarbons and to Remove Water from Natural Gas  

SciTech Connect

The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world high-pressure conditions is being conducted to convince industry users of the efficiency and reliability of the process. The system was designed and fabricated by Membrane Technology and Research, Inc. (MTR) and installed and operated at BP Amoco's Pascagoula, MS plant. The Gas Research Institute is partially supporting the field demonstration and BP-Amoco helped install the unit and provided onsite operators and utilities. The gas processed by the membrane system meets pipeline specifications for dewpoint and BTU value and can be delivered without further treatment to the pipeline. Based on data from prior membrane module tests, the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. During the course of this project, MTR has sold 11 commercial units related to the field test technology, and by the end of this demonstration project the process will be ready for broader commercialization. A route to commercialization has been developed during this project and involves collaboration with other companies already servicing the natural gas processing industry.

R. Baker; T. Hofmann; K. A. Lokhandwala

2005-09-29T23:59:59.000Z

230

Field Demonstration of a Membrane Process to Recover Heavy Hydrocarbons and to Remove Water from Natural Gas  

SciTech Connect

The objective of this project was to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world high-pressure conditions was conducted to convince industry users of the efficiency and reliability of the process. The system was designed and fabricated by Membrane Technology and Research, Inc. (MTR) and installed and operated at BP Amoco's Pascagoula, MS plant. The Gas Research Institute partially supported the field demonstration and BP-Amoco helped install the unit and provide onsite operators and utilities. The gas processed by the membrane system meets pipeline specifications for dew point and BTU value and can be delivered without further treatment to the pipeline. During the course of this project, MTR has sold thirteen commercial units related to the field test technology. Revenue generated from new business is already more than four times the research dollars invested in this process by DOE. The process is ready for broader commercialization and the expectation is to pursue the commercialization plans developed during this project, including collaboration with other companies already servicing the natural gas processing industry.

Kaaeid Lokhandwala

2007-03-30T23:59:59.000Z

231

Field Demonstration of a Membrane Process to Recover Heavy Hydrocarbons and to Remove Water from Natural Gas  

Science Conference Proceedings (OSTI)

The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world high-pressure conditions is being conducted to convince industry users of the efficiency and reliability of the process. The system was designed and fabricated by Membrane Technology and Research, Inc. (MTR) and installed and operated at BP Amoco's Pascagoula, MS plant. The Gas Research Institute is partially supporting the field demonstration and BP-Amoco helped install the unit and provides onsite operators and utilities. The gas processed by the membrane system meets pipeline specifications for dew point and BTU value and can be delivered without further treatment to the pipeline. Based on data from prior membrane module tests, the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. During the course of this project, MTR has sold 13 commercial units related to the field test technology, and by the end of this demonstration project the process will be ready for broader commercialization. A route to commercialization has been developed during this project and involves collaboration with other companies already servicing the natural gas processing industry.

R. Baker; T. Hofmann; K. A. Lokhandwala

2006-09-29T23:59:59.000Z

232

Magnetically assisted chemical separation (MACS) process: Preparation and optimization of particles for removal of transuranic elements  

SciTech Connect

The Magnetically Assisted Chemical Separation (MACS) process combines the selectivity afforded by solvent extractants with magnetic separation by using specially coated magnetic particles to provide a more efficient chemical separation of transuranic (TRU) elements, other radionuclides, and heavy metals from waste streams. Development of the MACS process uses chemical and physical techniques to elucidate the properties of particle coatings and the extent of radiolytic and chemical damage to the particles, and to optimize the stages of loading, extraction, and particle regeneration. This report describes the development of a separation process for TRU elements from various high-level waste streams. Polymer-coated ferromagnetic particles with an adsorbed layer of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) diluted with tributyl phosphate (TBP) were evaluated for use in the separation and recovery of americium and plutonium from nuclear waste solutions. Due to their chemical nature, these extractants selectively complex americium and plutonium contaminants onto the particles, which can then be recovered from the solution by using a magnet. The partition coefficients were larger than those expected based on liquid[liquid extractions, and the extraction proceeded with rapid kinetics. Extractants were stripped from the particles with alcohols and 400-fold volume reductions were achieved. Particles were more sensitive to acid hydrolysis than to radiolysis. Overall, the optimization of a suitable NMCS particle for TRU separation was achieved under simulant conditions, and a MACS unit is currently being designed for an in-lab demonstration.

Nunez, L.; Kaminski, M.; Bradley, C.; Buchholz, B.A.; Aase, S.B.; Tuazon, H.E.; Vandegrift, G.F. [Argonne National Lab., IL (United States); Landsberger, S. [Univ. of Illinois, Urbana, IL (United States)

1995-05-01T23:59:59.000Z

233

Actinide extraction from ICPP sodium bearing waste with 0.75 M DHDECMP/TBP in Isopar L{reg_sign}  

SciTech Connect

Recent process development efforts at the Idaho Chemical Processing Plant include examination of solvent extraction technologies for actinide partitioning from sodium bearing waste (SBW) solutions. The use of 0.75 {und M} dihexyl-N, N-diethylcarbamoylmethylphosphonate (DHDECMP or simply CMP) and 1.0 {und M} tri-n-butyl phosphate (TBP) diluted in Isopar L{reg_sign} was explored for actinide removal from simulated SBW solutions. Experimental evaluations included batch contacts in radiotracer tests with simulated sodium bearing waste solution to measure the extraction and recovery efficiency of the organic solvent. The radioactive isotopes utilized for this study included Pu-238, Pu-239, Am-241, U-233, Np-239, Zr-95, Tc-99m, and Hg-203. Extraction contacts of the organic solvent with the traced SBW stimulant, strip (back-extraction) contacts of the loaded organic solvent with either a 1-hydroxyethane-1, 1-diphosphonic acid (HEDPA) in nitric acid solution or an oxalic acid in nitric acid solution, and solvent wash contacts with sodium carbonate were performed.

Herbst, R.S.; Brewer, K.N.; Garn, T.G.; Law, J.D.; Rodriguez, A.M.; Tillotson, R.T.

1996-01-01T23:59:59.000Z

234

Ultratrace analysis of transuranic actinides by laser-induced fluorescence  

DOE Patents (OSTI)

Ultratrace quantities of transuranic actinides are detected indirectly by their effect on the fluorescent emissions of a preselected fluorescent species. Transuranic actinides in a sample are coprecipitated with a host lattice material containing at least one preselected fluorescent species. The actinide either quenches or enhances the laser-induced fluorescence of the preselected fluorescent species. The degree of enhancement or quenching is quantitatively related to the concentration of actinide in the sample.

Miller, Steven M. (Chelmsford, MA)

1988-01-01T23:59:59.000Z

235

Process Optimization for Solid Extraction, Flavor Improvement and Fat Removal in the Production of Soymilk From Full Fat Soy Flakes  

SciTech Connect

Traditionally soymilk has been made with whole soybeans; however, there are other alternative raw ingredients for making soymilk, such as soy flour or full-fat soy flakes. US markets prefer soymilk with little or no beany flavor. modifying the process or using lipoxygenase-free soybeans can be used to achieve this. Unlike the dairy industry, fat reduction in soymilk has been done through formula modification instead of by conventional fat removal (skimming). This project reports the process optimization for solids and protein extraction, flavor improvement and fat removal in the production of 5, 8 and 12 {sup o}Brix soymilk from full fat soy flakes and whole soybeans using the Takai soymilk machine. Proximate analyses, and color measurement were conducted in 5, 8 and 12 {sup o}Brix soymilk. Descriptive analyses with trained panelists (n = 9) were conducted using 8 and 12 {sup o}Brix lipoxygenase-free and high protein blend soy flake soymilks. Rehydration of soy flakes is necessary to prevent agglomeration during processing and increase extractability. As the rehydration temperature increases from 15 to 50 to 85 C, the hexanal concentration was reduced. Enzyme inactivation in soy flakes milk production (measured by hexanal levels) is similar to previous reports with whole soybeans milk production; however, shorter rehydration times can be achieved with soy flakes (5 to 10 minutes) compared to whole beans (8 to 12 hours). Optimum rehydration conditions for a 5, 8 and 12 {sup o}Brix soymilk are 50 C for 5 minutes, 85 C for 5 minutes and 85 C for 10 minutes, respectively. In the flavor improvement study of soymilk, the hexanal date showed differences between undeodorized HPSF in contrast to triple null soymilk and no differences between deodorized HPSF in contrast to deodorized triple null. The panelists could not differentiate between the beany, cereal, and painty flavors. However, the panelists responded that the overall aroma of deodorized 8 {sup o}Brix triple null and HPSF soymilk are lower than the undeodorized triple null and HPSF soymilk. The triple null soymilk was perceived to be more bitter than the HPSF soymilk by the sensory panel due to oxidation on the triple null soy flakes. This oxidation may produce other aroma that was not analyzed using the GC but noticed by the panelists. The sensory evaluation results did show that the deodorizer was able to reduce the soymilk aroma in HPSF soymilk so it would be similar to triple null soymilk at 8 {sup o}Brix level. Regardless of skimming method and solids levels, the fat from the whole soybean milk was removed less efficiently than soy flake milk (7 to 30% fat extraction in contrast to 50 to 80% fat extraction respectively). In soy flake milk, less fat was removed as the % solid increases regardless of the processing method. In whole soybean milk, the fat was removed less efficiently at lower solids level milk using the commercial dairy skimmer and more efficient at lower solids level using the centrifuge-decant method. Based on the Hunter L, a, b measurement, the color of the reduced fat soy flake milk yielded a darker, greener and less yellow colored milk than whole soymilk ({alpha} < 0.05), whereas no differences were noticed in reduced fat soybean milk ({alpha} < 0.05). Color comparison of whole and skim cow's milk showed the same the same trend as in the soymilk.

Stanley Prawiradjaja

2003-05-31T23:59:59.000Z

236

SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS  

DOE Patents (OSTI)

A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

Schubert, J.

1960-05-24T23:59:59.000Z

237

Process for removal of mineral particulates from coal-derived liquids  

SciTech Connect

Suspended mineral solids are separated from a coal-derived liquid containing the solids by a process comprising the steps of: (a) contacting said coal-derived liquid containing solids with a molten additive having a melting point of 100.degree.-500.degree. C. in an amount of up to 50 wt. % with respect to said coal-derived liquid containing solids, said solids present in an amount effective to increase the particle size of said mineral solids and comprising material or mixtures of material selected from the group of alkali metal hydroxides and inorganic salts having antimony, tin, lithium, sodium, potassium, magnesium, calcium, beryllium, aluminum, zinc, molybdenum, cobalt, nickel, ruthenium, rhodium or iron cations and chloride, iodide, bromide, sulfate, phosphate, borate, carbonate, sulfite, or silicate anions; and (b) maintaining said coal-derived liquid in contact with said molten additive for sufficient time to permit said mineral matter to agglomerate, thereby increasing the mean particle size of said mineral solids; and (c) recovering a coal-derived liquid product having reduced mineral solids content. The process can be carried out with less than 5 wt. % additive and in the absence of hydrogen pressure.

McDowell, William J. (Knoxville, TN)

1980-01-01T23:59:59.000Z

238

Fission theory and actinide fission data  

SciTech Connect

The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and ..gamma.. rays emitted in the resonances of /sup 235/U and /sup 239/Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and ..gamma..-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n, ..gamma.., f) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for /sup 240/Pu. 17 figures, 56 ref. (auth)

Michaudon, A.

1975-10-01T23:59:59.000Z

239

HIGH TEMPERATURE REMOVAL OF H{sub 2}S FROM COAL GASIFICATION PROCESS STREAMS USING AN ELECTROCHEMICAL MEMBRANE SYSTEM  

SciTech Connect

A bench scale set-up was constructed to test the cell performance at 600-700 C and 1 atm. The typical fuel stream inlet proportions were 34% CO, 22% CO{sub 2}, 35% H{sub 2}, 8% H{sub 2}O, and 450-2000 ppm H{sub 2}S. The fundamental transport restrictions for sulfur species in an electrochemical cell were examined. Temperature and membrane thickness were varied to examine how these parameters affect the maximum flux of H{sub 2}S removal. It was found that higher temperature allows more sulfide species to enter the electrolyte, thus increasing the sulfide flux across the membrane and raising the maximum flux of H{sub 2}S removal. The results identify sulfide diffusion across the membrane as the rate-limiting step in H{sub 2}S removal. The maximum H{sub 2}S removal flux of 1.1 x 10-6 gmol H{sub 2}S min{sup -1} cm{sup -2} (or 3.5 mA cm{sup -2}) was obtained at 650 C, with a membrane that was 0.9 mm thick, 36% porous, and had an estimated tortuosity of 3.6. Another focus of this thesis was to examine the stability of cathode materials in full cell trials. A major hurdle that remains in process scale-up is cathode selection, as the lifetime of the cell will depend heavily on the lifetime of the cathode material, which is exposed to very sour gas. Materials that showed success in the past (i.e. cobalt sulfides and Y{sub 0.9}Ca{sub 0.1}FeO{sub 3}) were examined but were seen to have limitations in operating environment and temperature. Therefore, other novel metal oxide compounds were studied to find possible candidates for full cell trials. Gd{sub 2}TiMoO{sub 7} and La{sub 0.7}Sr{sub 0.3}VO{sub 3} were the compounds that retained their structure best even when exposed to high H{sub 2}S, CO{sub 2}, and H{sub 2}O concentrations.

Jack Winnick; Meilin Liu

2003-06-01T23:59:59.000Z

240

Operation of a bushing melter system designed for actinide vitrification  

SciTech Connect

The Westinghouse Savannah River Company is developing a melter system to vitrify actinide materials. The melter system will used to vitrify the americium and curium solution which is currently stored in one of the Savannah River Site`s (SRS) processing canyons. This solution is one of the materials designated by the Defense Nuclear Facilities Safety Board (DNFSB) to be dispositioned as part of the DNFSB recommendation 94-1. The Am/Cm solution contains an extremely large fraction (>2 kilograms of Cm and 10 kilograms of Am) of t he United States`s total inventory of both elements. They have an estimated value on the order of one billion dollars - if they are processed through the DOE Isotope Sales program at the Oak Ridge National Laboratory. It is therefore deemed highly desirable to transfer the material to Oak Ridge in a form which can allow for recovery of the material. A commercial glass composition has been demonstrated to be compatible with up to 40 weight percent of the Am/Cm solution contents. This glass is also selectively attacked by nitric acid. This allows the actinide to be recovered by common separation processes.

Ramsey, W.G.

1996-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Process studies for a new method of removing H/sub 2/S from industrial gas streams  

SciTech Connect

A process for the removal of hydrogen sulfide from coal-derived gas streams has been developed. The basis for the process is the absorption of H/sub 2/S into a polar organic solvent where it is reacted with dissolved sulfur dioxide to form elemental sulfur. After sulfur is crystallized from solution, the solvent is stripped to remove dissolved gases and water formed by the reaction. The SO/sub 2/ is generated by burning a portion of the sulfur in a furnace where the heat of combustion is used to generate high pressure steam. The SO/sub 2/ is absorbed into part of the lean solvent to form the solution necessary for the first step. The kinetics of the reaction between H/sub 2/S and SO/sub 2/ dissolved in mixtures of N,N-Dimethylaniline (DMA)/ Diethylene Glycol Monomethyl Ether and DMA/Triethylene Glycol Dimethyl Ether was studied by following the temperature rise in an adiabatic calorimeter. This irreversible reaction was found to be first-order in both H/sub 2/S and SO/sub 2/, with an approximates heat of reaction of 28 kcal/mole of SO/sub 2/. The sole products of the reaction appear to be elemental sulfur and water. The presence of DMA increases the value of the second-order rate constant by an order of magnitude over that obtained in the glycol ethers alone. Addition of other tertiary aromatic amines enhances the observed kinetics; heterocyclic amines (e.g., pyridine derivatives) have been found to be 10 to 100 times more effective as catalysts when compared to DMA.

Neumann, D.W.; Lynn, S.

1986-07-01T23:59:59.000Z

242

Process for simultaneous removal of SO.sub.2 and NO.sub.x from gas streams  

SciTech Connect

A process for simultaneous removal of SO.sub.2 and NO.sub.x from a gas stream that includes flowing the gas stream to a spray dryer and absorbing a portion of the SO.sub.2 content of the gas stream and a portion of the NO.sub.x content of the gas stream with ZnO by contacting the gas stream with a spray of an aqueous ZnO slurry; controlling the gas outlet temperature of the spray dryer to within the range of about a 0.degree. to 125.degree. F. approach to the adiabatic saturation temperature; flowing the gas, unreacted ZnO and absorbed SO.sub.2 and NO.sub.x from the spray dryer to a fabric filter and collecting any solids therein and absorbing a portion of the SO.sub.2 remaining in the gas stream and a portion of the NO.sub.x remaining in the gas stream with ZnO; and controlling the ZnO content of the aqueous slurry so that sufficient unreacted ZnO is present in the solids collected in the fabric filter to react with SO.sub.2 and NO.sub.x as the gas passes through the fabric filter whereby the overall feed ratio of ZnO to SO.sub.2 plus NO.sub.x is about 1.0 to 4.0 moles of ZnO per of SO.sub.2 and about 0.5 to 2.0 moles of ZnO per mole of NO.sub.x. Particulates may be removed from the gas stream prior to treatment in the spray dryer. The process further allows regeneration of ZnO that has reacted to absorb SO.sub.2 and NO.sub.x from the gas stream and acid recovery.

Rosenberg, Harvey S. (Columbus, OH)

1987-01-01T23:59:59.000Z

243

Process for simultaneous removal of SO[sub 2] and NO[sub x] from gas streams  

DOE Patents (OSTI)

A process is described for simultaneous removal of SO[sub 2] and NO[sub x] from a gas stream that includes flowing the gas stream to a spray dryer and absorbing a portion of the SO[sub 2] content of the gas stream and a portion of the NO[sub x] content of the gas stream with ZnO by contacting the gas stream with a spray of an aqueous ZnO slurry; controlling the gas outlet temperature of the spray dryer to within the range of about a 0 to 125 F approach to the adiabatic saturation temperature; flowing the gas, unreacted ZnO and absorbed SO[sub 2] and NO[sub x] from the spray dryer to a fabric filter and collecting any solids therein and absorbing a portion of the SO[sub 2] remaining in the gas stream and a portion of the NO[sub x] remaining in the gas stream with ZnO; and controlling the ZnO content of the aqueous slurry so that sufficient unreacted ZnO is present in the solids collected in the fabric filter to react with SO[sub 2] and NO[sub x] as the gas passes through the fabric filter whereby the overall feed ratio of ZnO to SO[sub 2] plus NO[sub x] is about 1.0 to 4.0 moles of ZnO per of SO[sub 2] and about 0.5 to 2.0 moles of ZnO per mole of NO[sub x]. Particulates may be removed from the gas stream prior to treatment in the spray dryer. The process further allows regeneration of ZnO that has reacted to absorb SO[sub 2] and NO[sub x] from the gas stream and acid recovery. 4 figs.

Rosenberg, H.S.

1987-02-03T23:59:59.000Z

244

Demonstration of the SREX process for the removal of {sup 90}Sr from actual highly radioactive solutions in centrifugal contactors  

Science Conference Proceedings (OSTI)

The SREX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of {sup 90}Sr from acidic radioactive wastes stored at the ICPP. These efforts have culminated in a recent demonstration of the SREX process with actual tank waste. This demonstration was performed using 24 stages of 2-cm diameter centrifugal contactors installed in a shielded hot cell at the ICPP Remote Analytical Laboratory. An overall removal efficiency of 99.995% was obtained for {sup 90}Sr. As a result, the activity of {sup 90}Sr was reduced from 201 Ci/m{sup 3} in the feed solution of 0.0089 Ci/m{sup 3} in the aqueous raffinate, which is below the U.S. NRC Class A LLW limit of 0.04 Ci/m{sup 3} for {sup 90}Sr. Lead was extracted by the SREX solvent and successfully partitioned from the {sup 90}Sr using an ammonium citrate strip solution. Additionally, 94% of the total alpha activity, 1.9% of the {sup 241}Am, 99.94% of the {sup 238}Pu, 99.97% of the {sup 239}Pu, 36.4% of the K, 64% of the Ba, and >83% of the Zr were extracted by the SREX solvent. Cs, B, Cd, Ca, Cr, Fe, Mn, Ni, and Na were essentially inextractable. 10 refs., 2 figs., 3 tabs.

Law, J.D.; Wood, D.J.; Todd, T.A.; Olson, L.G.

1997-10-01T23:59:59.000Z

245

Process for removing halogenated aliphatic and aromatic compounds from petroleum products. [Polychlorinated biphenyls; methylene chloride; perchloroethylene; trichlorofluoroethane; trichloroethylene; chlorobenzene  

DOE Patents (OSTI)

A process for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contracting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible polyhydroxy compound, such as, water, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds in the low polar or nonpolar solvent by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered for recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced. 2 tables.

Googin, J.M.; Napier, J.M.; Travaglini, M.A.

1982-03-31T23:59:59.000Z

246

Actinide consumption: Nuclear resource conservation without breeding  

SciTech Connect

A new approach to the nuclear power issue based on a metallic fast reactor fuel and pyrometallurgical processing of spent fuel is showing great potential and is approaching a critical demonstration phase. If successful, this approach will complement and validate the LWR reactor systems and the attendant infrastructure (including repository development) and will alleviate the dominant concerns over the acceptability of nuclear power. The Integral Fast Reactor (IFR) concept is a metal-fueled, sodium-cooled pool-type fast reactor supported by a pyrometallurgical reprocessing system. The concept of a sodium cooled fast reactor is broadly demonstrated by the EBR-II and FFTF in the US; DFR and PFR in the UK; Phenix and SuperPhenix in France; BOR-60, BN-350, BN-600 in the USSR; and JOYO in Japan. The metallic fuel is an evolution from early EBR-II fuels. This fuel, a ternary U-Pu-Zr alloy, has been demonstrated to be highly reliable and fault tolerant even at very high burnup (160-180,000 MWd/MT). The fuel, coupled with the pool type reactor configuration, has been shown to have outstanding safety characteristics: even with all active safety systems disabled, such a reactor can survive a loss of coolant flow, a loss of heat sink, or other major accidents. Design studies based on a small modular approach show not only its impressive safety characteristics, but are projected to be economically competitive. The program to explore the feasibility of actinide recovery from spent LWR fuel is in its initial phase, but it is expected that technical feasibility could be demonstrated by about 1995; DOE has not yet committed funds to achieve this objective. 27 refs.

Hannum, W.H.; Battles, J.E.; Johnson, T.R.; McPheeters, C.C.

1991-01-01T23:59:59.000Z

247

URANIUM SEPARATION PROCESS  

DOE Patents (OSTI)

A method of separating uranium oxides from PuO/sub 2/, ThO/sub 2/, and other actinide oxides is described. The oxide mixture is suspended in a fused salt melt and a chlorinating agent such as chlorine gas or phosgene is sparged through the suspension. Uranium oxides are selectively chlorinated and dissolve in the melt, which may then be filtered to remove the unchlorinated oxides of the other actinides. (AEC)

Lyon, W.L.

1962-04-17T23:59:59.000Z

248

Silica Scaling Removal Process  

NLE Websites -- All DOE Office Websites (Extended Search)

systems Water treatment systems Water evaporation systems Potential mining applications (produced water) Industry applications for which silica scaling must be prevented Benefits:...

249

Silica Scaling Removal Process  

NLE Websites -- All DOE Office Websites (Extended Search)

applications for which silica scaling must be prevented Benefits: Reduces scaling in cooling towers by up to 50% Increases the number of cycles of concentration substantially...

250

Comparative evaluation of DHDECMP (dihexyl-N,N-diethylcarbamoyl-methylphosphonate) and CMPO (octylphenyl-N,N,-diisobutylcarbamoylmethylphosphine oxide) as extractants for recovering actinides from nitric acid waste streams  

SciTech Connect

Certain neutral, bifunctional organophosphorous compounds are of special value to the nuclear industry. Dihexyl-N,N-diethylcarbomoylmethylphosphonate (DHDECMP) and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) are highly selective extractants for removing actinide and lanthanide elements from nitric acid. We obtained these two extractants from newly available commercial sources and evaluated them for recovering Am(III), Pu(IV), and U(VI) from nitric acid waste streams of plutonium processing operations. Variables included the extractant (DHSECMP or CMPO), extractant/tributylphosphate ratio, diluent, nitrate concentration, nitrate salt/nitric acid ratio, fluoride concentration, and contact time. Based on these experimental data, we selected DHDECMP as the perferred extractant for this application. 18 refs., 30 figs.

Marsh, S.F.; Yarbro, S.L.

1988-02-01T23:59:59.000Z

251

Low Cost Chemical Feedstocks Using an Improved and Energy Efficient Natural Gas Liquid (NGL) Removal Process, Final Technical Report  

SciTech Connect

The overall objective of this project is to develop a new low-cost and energy efficient Natural Gas Liquid (NGL) recovery process - through a combination of theoretical, bench-scale and pilot-scale testing - so that it could be offered to the natural gas industry for commercialization. The new process, known as the IROA process, is based on U.S. patent No. 6,553,784, which if commercialized, has the potential of achieving substantial energy savings compared to currently used cryogenic technology. When successfully developed, this technology will benefit the petrochemical industry, which uses NGL as feedstocks, and will also benefit other chemical industries that utilize gas-liquid separation and distillation under similar operating conditions. Specific goals and objectives of the overall program include: (i) collecting relevant physical property and Vapor Liquid Equilibrium (VLE) data for the design and evaluation of the new technology, (ii) solving critical R&D issues including the identification of suitable dehydration and NGL absorbing solvents, inhibiting corrosion, and specifying proper packing structure and materials, (iii) designing, construction and operation of bench and pilot-scale units to verify design performance, (iv) computer simulation of the process using commercial software simulation platforms such as Aspen-Plus and HYSYS, and (v) preparation of a commercialization plan and identification of industrial partners that are interested in utilizing the new technology. NGL is a collective term for C2+ hydrocarbons present in the natural gas. Historically, the commercial value of the separated NGL components has been greater than the thermal value of these liquids in the gas. The revenue derived from extracting NGLs is crucial to ensuring the overall profitability of the domestic natural gas production industry and therefore of ensuring a secure and reliable supply in the 48 contiguous states. However, rising natural gas prices have dramatically reduced the economic incentive to extract NGLs from domestically produced natural gas. Successful gas processors will be those who adopt technologies that are less energy intensive, have lower capital and operating costs and offer the flexibility to tailor the plant performance to maximize product revenue as market conditions change, while maintaining overall system efficiency. Presently, cryogenic turbo-expander technology is the dominant NGL recovery process and it is used throughout the world. This process is known to be highly energy intensive, as substantial energy is required to recompress the processed gas back to pipeline pressure. The purpose of this project is to develop a new NGL separation process that is flexible in terms of ethane rejection and can reduce energy consumption by 20-30% from current levels, particularly for ethane recoveries of less than 70%. The new process integrates the dehydration of the raw natural gas stream and the removal of NGLs in such a way that heat recovery is maximized and pressure losses are minimized so that high-value equipment such as the compressor, turbo-expander, and a separate dehydration unit are not required. GTI completed a techno-economic evaluation of the new process based on an Aspen-HYSYS simulation model. The evaluation incorporated purchased equipment cost estimates obtained from equipment suppliers and two different commercial software packages; namely, Aspen-Icarus and Preliminary Design and Quoting Service (PDQ$). For a 100 MMscfd gas processing plant, the annualized capital cost for the new technology was found to be about 10% lower than that of conventional technology for C2 recovery above 70% and about 40% lower than that of conventional technology for C2 recovery below 50%. It was also found that at around 40-50% C2 recovery (which is economically justifiable at the current natural gas prices), the energy cost to recover NGL using the new technology is about 50% of that of conventional cryogenic technology.

Meyer, Howard, S.; Lu, Yingzhong

2012-08-10T23:59:59.000Z

252

The role of actinide burning and the Integral Fast Reactor in the future of nuclear power  

Science Conference Proceedings (OSTI)

A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

1990-12-01T23:59:59.000Z

253

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

Science Conference Proceedings (OSTI)

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 a 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.

Jackson, Jay M. [Los Alamos National Laboratory; Lopez, Jacquelyn C. [Los Alamos National Laboratory; Wayne, David M. [Los Alamos National Laboratory; Schulte, Louis D. [Los Alamos National Laboratory; Finstad, Casey C. [Los Alamos National Laboratory; Stroud, Mary Ann [Los Alamos National Laboratory; Mulford, Roberta Nancy [Los Alamos National Laboratory; MacDonald, John M. [Los Alamos National Laboratory; Turner, Cameron J. [Los Alamos National Laboratory; Lee, Sonya M. [Los Alamos National Laboratory

2012-07-05T23:59:59.000Z

254

Effect of residual chips on the material removal process of the bulk metallic glass studied by in situ scratch testing inside the scanning electron microscope  

Science Conference Proceedings (OSTI)

Research on material removal mechanism is meaningful for precision and ultra-precision manufacturing. In this paper, a novel scratch device was proposed by integrating the parasitic motion principle linear actuator. The device has a compact structure and it can be installed on the stage of the scanning electron microscope (SEM) to carry out in situ scratch testing. Effect of residual chips on the material removal process of the bulk metallic glass (BMG) was studied by in situ scratch testing inside the SEM. The whole removal process of the BMG during the scratch was captured in real time. Formation and growth of lamellar chips on the rake face of the Cube-Corner indenter were observed dynamically. Experimental results indicate that when lots of chips are accumulated on the rake face of the indenter and obstruct forward flow of materials, materials will flow laterally and downward to find new location and direction for formation of new chips. Due to similar material removal processes, in situ scratch testing is potential to be a powerful research tool for studying material removal mechanism of single point diamond turning, single grit grinding, mechanical polishing and grating fabrication.

Huang Hu; Zhao Hongwei; Shi Chengli; Wu Boda; Fan Zunqiang; Wan Shunguang; Geng Chunyang [College of Mechanical Science and Engineering, Jilin University, Renmin Street 5988, Changchun, Jilin 130025 (China)

2012-12-15T23:59:59.000Z

255

Method for decontamination of nickel-fluoride-coated nickel containing actinide-metal fluorides  

DOE Patents (OSTI)

The invention is a process for decontaminating particulate nickel contaminated with actinide-metal fluorides. In one aspect, the invention comprises contacting nickel-fluoride-coated nickel with gaseous ammonia at a temperature effecting nickel-catalyzed dissociation thereof and effecting hydrogen-reduction of the nickel fluoride. The resulting nickel is heated to form a melt and a slag and to effect transfer of actinide metals from the melt into the slag. The melt and slag are then separated. In another aspect, nickel containing nickel oxide and actinide metals is contacted with ammonia at a temperature effecting nickel-catalyzed dissociation to effect conversion of the nickel oxide to the metal. The resulting nickel is then melted and separated as described. In another aspect nickel-fluoride-coated nickel containing actinide-metal fluorides is contacted with both steam and ammonia. The resulting nickel then is melted and separated as described. The invention is characterized by higher nickel recovery, efficient use of ammonia, a substantial decrease in slag formation and fuming, and a valuable increase in the service life of the furnace liners used for melting.

Windt, Norman F. (Paducah, KY); Williams, Joe L. (Paducah, KY)

1983-01-01T23:59:59.000Z

256

Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide as an extractant for actinides from nitric acid waste  

SciTech Connect

The ability of neutral bifunctional organophosphorus compounds to extract trivalent actinides, specifically Am(III), from nitric acid solution has stimulated investigations into the processing of PUREX high level liquid waste. The authors' work in this area has focused primarily on derivatives of carbamoylmethylphosphine oxide (CMPO). The authors of this paper have found that the best extractant of this type is octyl(phenyl)-N,N-diisobutylmethylphosphine oxide (abbrev. O phi D (IB)CMPO). This extractant has a unique combination of substituent groups which impart to the resulting molecule substantially improved ability to extract actinides from acidic nitrate media and to withstand hydrolytic and radiolytic degradation. At the same time good selectivity of Am(III) actinides over fission products and favorable solubility properties on actinide loading are maintained. This paper describes the application of O phi D (IB) CMPO to the extraction of transuranium elements (Np, Pu, and Am) and fission product rare earths (F.P.R.E.) from evaporated highly acidic (5 M HNO/sub 3/) PUREX waste. Additional information on the influence of phenyl substitution in CMPO's and mixed CMPO-TBP extractant solutions on D/sub Am/ is also presented. 3 figures, 2 tables.

Horwitz, E.P.; Diamond, H.; Kalina, D.G.; Kaplan, L.; Mason, G.W.

1983-01-01T23:59:59.000Z

257

Surrogate Reactions in the Actinide Region  

SciTech Connect

Over the past three years we have studied various surrogate reactions (d,p), ({sup 3}He,t), ({alpha},{alpha}{sup '}) on several uranium isotopes {sup 234}U, {sup 235}U, {sup 236}U, and {sup 238}U. An overview of the STARS/LIBERACE surrogate research program as it pertains to the actinides is discussed. A summary of results to date will be presented along with a discussion of experimental difficulties encountered in surrogate experiments and future research directions.

Burke, J. T.; Bernstein, L. A.; Scielzo, N. D.; Bleuel, D. L.; Lesher, S. R.; Escher, J.; Ahle, L.; Dietrich, F. S.; Hoffman, R. D.; Norman, E. B.; Sheets, S. A.; Phair, L.; Fallon, P.; Clark, R. M.; Gibelin, J.; Jewett, C.; Lee, I. Y.; Macchiavelli, A. O.; McMahan, M. A.; Moretto, L. G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Lawrence Berkeley National Laboratory, Berkeley, California, 94720 (United States); University of California, Berkeley, Berkeley, California, 94720 (United States); University of Richmond, Richmond, Virginia, 23173 (United States); Yale University, New Haven, Connecticut, 06520 (United States); Rutgers University, New Brunswick, New Jersey, 08901 (United States)] (and others)

2008-04-17T23:59:59.000Z

258

Joint Actinide Shock Physics Experimental Research Facility Restart...  

NLE Websites -- All DOE Office Websites (Extended Search)

Office of Safety and Emergency Management Evaluations Activity Report for the Joint Actinide Shock Physics Experimental Research Facility Restart Operational Readiness...

259

Complexation of Actinides in Solution: Thermodynamic Measurements and Structural Characterization  

E-Print Network (OSTI)

of Actinides in Solution: Thermodynamic Measurements andAn integrated approach of thermodynamic measurements andAn integrated approach of thermodynamic measurements and

Rao, L.

2007-01-01T23:59:59.000Z

260

REPORT: Inert-Matrix Fuel: Actinide ''Burning'' and Direct ... - TMS  

Science Conference Proceedings (OSTI)

Jun 27, 2007 ... Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am ...

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
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261

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Albright, D. , Plutonium and highly enriched uranium, 1996 :and swelling in uranium-plutonium mixed nitride fuels.products and to extract plutonium or any other actinide from

Heidet, Florent

2010-01-01T23:59:59.000Z

262

Advanced Nuclear Fuel Concepts for Minor Actinide Burning  

Science Conference Proceedings (OSTI)

Abstract Scope, New fuel cycle strategies entail advanced nuclear fuel concepts. This especially applies for the burning of minor actinides in a fast reactor cycle ...

263

Joint Actinide Shock Physics Experimental Research Facility Restart...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

and Emergency Management Evaluations Activity Report for the Joint Actinide Shock Physics Experimental Research Facility Restart Operational Readiness Review Pre- Visit Dates...

264

Evaluation of an alkaline-side solvent extraction process for cesium removal from SRS tank waste using laboratory-scale centrifugal contactors  

Science Conference Proceedings (OSTI)

An alkaline-side solvent extraction process for cesium removal from Savannah River Site (SRS) tank waste was evaluated experimentally using a laboratory-scale centrifugal contactor. Single-stage and multistage tests were conducted with this contactor to determine hydraulic performance, stage efficiency, and general operability of the process flowsheet. The results and conclusions of these tests are reported along with those from various supporting tests. Also discussed is the ability to scale-up from laboratory- to plant-scale operation when centrifugal contractors are used to carry out the solvent extraction process. While some problems were encountered, a promising solution for each problem has been identified. Overall, this alkaline-side cesium extraction process appears to be an excellent candidate for removing cesium from SRS tank waste.

Leonard, R. A.; Conner, C.; Liberatore, M. W.; Sedlet, J.; Aase, S. B.; Vandegrift, G. F.

1999-11-29T23:59:59.000Z

265

Development of an Integrated Multi-Contaminant Removal Process Applied to Warm Syngas Cleanup for Coal-Based Advanced Gasification Systems  

Science Conference Proceedings (OSTI)

This project met the objective to further the development of an integrated multi-contaminant removal process in which H2S, NH3, HCl and heavy metals including Hg, As, Se and Cd present in the coal-derived syngas can be removed to specified levels in a single/integrated process step. The process supports the mission and goals of the Department of Energyâ??s Gasification Technologies Program, namely to enhance the performance of gasification systems, thus enabling U.S. industry to improve the competitiveness of gasification-based processes. The gasification program will reduce equipment costs, improve process environmental performance, and increase process reliability and flexibility. Two sulfur conversion concepts were tested in the laboratory under this project, i.e., the solventbased, high-pressure University of California Sulfur Recovery Process â?? High Pressure (UCSRP-HP) and the catalytic-based, direct oxidation (DO) section of the CrystaSulf-DO process. Each process required a polishing unit to meet the ultra-clean sulfur content goals of <50 ppbv (parts per billion by volume) as may be necessary for fuel cells or chemical production applications. UCSRP-HP was also tested for the removal of trace, non-sulfur contaminants, including ammonia, hydrogen chloride, and heavy metals. A bench-scale unit was commissioned and limited testing was performed with simulated syngas. Aspen-Plus®-based computer simulation models were prepared and the economics of the UCSRP-HP and CrystaSulf-DO processes were evaluated for a nominal 500 MWe, coal-based, IGCC power plant with carbon capture. This report covers the progress on the UCSRP-HP technology development and the CrystaSulf-DO technology.

Howard Meyer

2010-11-30T23:59:59.000Z

266

Supercritical Carbon Dioxide Ligands for Extracting Actinide Metal Ions from Porous Solids  

SciTech Connect

Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effective method for the removal of radioactive contaminants would release this valuable material for salvage.

Albert W. Herlinger; Dr. Mark L. Dietz

2003-03-06T23:59:59.000Z

267

Supercritical Carbon Dioxide-Soluble Ligands for Extracting Actinide Metal Ions from Porous Solids  

SciTech Connect

Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effect method for the removal of radioactive contaminants would release this valuable material for salvage.

Joan Brennecke; Mark Dietz; Richard Barrans; Alabert Herlinger

2003-07-03T23:59:59.000Z

268

POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS  

Science Conference Proceedings (OSTI)

A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

PUIGH RJ; TOFFER H

2011-10-19T23:59:59.000Z

269

Method for extracting lanthanides and actinides from acid solutions by modification of Purex solvent  

DOE Patents (OSTI)

A process is described for the recovery of actinide and lanthanide values from aqueous solutions with an extraction solution containing an organic extractant having the formula as shown in a diagram where [phi] is phenyl, R[sup 1] is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R[sup 2] is an alkyl containing from 3 to 6 carbon atoms and phase modifiers in a water-immiscible hydrocarbon diluent. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions. 6 figs.

Horwitz, E.P.; Kalina, D.G.

1986-03-04T23:59:59.000Z

270

Method for extracting lanthanides and actinides from acid solutions by modification of purex solvent  

SciTech Connect

A process for the recovery of actinide and lanthanide values from aqueous solutions with an extraction solution containing an organic extractant having the formula: ##STR1## where .phi. is phenyl, R.sup.1 is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R.sup.2 is an alkyl containing from 3 to 6 carbon atoms and phase modifiers in a water-immiscible hydrocarbon diluent. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions.

Horwitz, E. Philip (Naperville, IL); Kalina, Dale G. (Naperville, IL)

1986-01-01T23:59:59.000Z

271

Evaluation of extractants and chelating resins in polishing actinide-contaminated waste streams  

SciTech Connect

At the Los Alamos National Laboratory Plutonium Facility, anion exchange is used for recovering plutonium from nitric acid solutions. Although this approach recovers >99%, the trace amounts of plutonium and other actinides remaining in the effluent require additional processing. We are doing research to develop a secondary unit operation that can directly polish the effluent so that actinide levels are reduced to below the maximum allowed for facility discharge. We selected solvent extraction, the only unit operation that can meet the stringent process requirements imposed; several carbonyl and phosphoryl extractants were evaluated and their performance characterized. We also investigated various engineering approaches for solvent extraction; the most promising was a chelating resin loaded with extractant. Our research now focuses on the synthesis of malonamides, and our goal is to bond these extractants to a resin matrix. 7 refs., 12 figs., 1 tab.

Schreiber, S.B.; Dunn, S.L.; Yarbro, S.L.

1991-06-01T23:59:59.000Z

272

METHOD FOR THE PREPARATION OF STABLE ACTINIDE METAL OXIDE-CONTAINING SLURRIES AND OF THE OXIDES THEREFOR  

DOE Patents (OSTI)

This patent deals with a method of preparing actinide metal oxides of a very fine particle size and of forming stable suspensions therefrom. The process consists of dissolving the nitrate of the actinide element in a combustible organic solvent, converting the solution obtained into a spray, and igniting the spray whereby an oxide powder is obtained. The oxide powder is then slurried in an aqueous soiution of a substance which is adsorbable by said oxides, dspersed in a colloid mill whereby a suspension is obtained, and electrodialyzed until a low spectiic conductance is reached.

Hansen, R.S.; Minturn, R.E.

1958-02-25T23:59:59.000Z

273

Progress toward Biomass and Coal-Derived Syngas Warm Cleanup: Proof-of-Concept Process Demonstration of Multicontaminant Removal for Biomass Application  

Science Conference Proceedings (OSTI)

Systems comprising of multiple sorbent and catalytic beds have been developed for the warm syngas cleanup of coal- and biomass-derived syngas. Tailored specifically for biomass application the process described here consists of six primary unit operations: 1) Na2CO3 bed for HCl removal, 2) two regenerable ZnO beds for bulk H2S removal, 3) ZnO bed for H2S polishing, 4) NiCu/SBA-16 sorbent for trace metal (e.g. AsH3) removal, 5) steam reforming catalyst bed for tars and light hydrocarbons reformation and NH3 decomposition, and a 6) Cu-based LT-WGS catalyst bed. Simulated biomass-derived syngas containing a multitude of inorganic contaminants (H2S, AsH3, HCl, and NH3) and hydrocarbon additives (methane, ethylene, benzene, and naphthalene) was used to demonstrate process effectiveness. The efficiency of the process was demonstrated for a period of 175 hours, during which no signs of deactivation were observed. Post-run analysis revealed small levels of sulfur slipped through the sorbent bed train to the two downstream catalytic beds. Future improvements could be made to the trace metal polishing sorbent to ensure complete inorganic contaminant removal (to low ppb level) prior to the catalytic steps. However, dual, regenerating ZnO beds were effective for continuous removal for the vast majority of the sulfur present in the feed gas. The process was effective for complete AsH3 and HCl removal. The steam reforming catalyst completely reformed all the hydrocarbons present in the feed (methane, ethylene, benzene, and naphthalene) to additional syngas. However, post-run evaluation, under kinetically-controlled conditions, indicates deactivation of the steam reforming catalyst. Spent material characterization suggests this is attributed, in part, to coke formation, likely due to the presence of benzene and/or naphthalene in the feed. Future adaptation of this technology may require dual, regenerable steam reformers. The process and materials described in this report hold promise for a warm cleanup of a variety of contaminant species within warm syngas.

Howard, Christopher J.; Dagle, Robert A.; Lebarbier, Vanessa MC; Rainbolt, James E.; Li, Liyu; King, David L.

2013-06-19T23:59:59.000Z

274

Downstream Processing of Recombinant Proteins from Transgenic Plant Systems: Phenolic Compounds Removal from Monoclonal Antibody Expressing Lemna minor and Purification of Recombinant Bovine Lysozyme from Sugarcane  

E-Print Network (OSTI)

Transgenic plant systems have been proposed as bioreactors in the production of pharmaceutical and industrial proteins. The economic benefits of inexpensive plant production systems could be erased if the downstream processing ends up being expensive. To avoid monoclonal antibody (mAb) modification or fouling of chromatography resins, removal of phenolics from plant extracts is desirable. Removal of major phenolics in Lemna extracts was evaluated by adsorption to PVPP, XAD-4, IRA-402 and Q-Sepharose resins. Analysis of phenolics adsorption to XAD-4, IRA-402 and Q-Sepharose showed superior dynamic binding capacities at pH 4.5 than at 7.5. The economic analysis using SuperPro Designer 7.0 indicated that addition of a phenolics adsorption step would increase mAb production cost only 20% by using IRA-402 compared to 35% for XAD-4 resin. The overall mAb processing cost can be reduced by implementing a phenolics removal step. To understand phenolics-resin interactions, adsorption isotherms of phenolic compounds (chlorogenic acid, ferulic acid, rutin, syringic acid and vitexin-2-O-rhamnoside) from different phenolic classes on three resins (IRA-402, PVPP, XAD-4) at pH 4.5 and 7.5 were determined. Differences in adsorption with the type of phenolics were observed, and PVPP was not efficient for phenolics removal. Screening of sugarcane lines for bovine lysozyme (BvLz) accumulation indicated that expression levels are still inadequate for commercial development. To maximize BvLz extraction, pH and ionic strength were evaluated; five conditions resulted in equivalent BvLz/TSP ratio. Membrane filtration process using BvLz extracts attained partial removal of native proteins by the 100 kDa membrane step, but also BvLz loss (21-29%). Regardless of the extraction condition, at least 47% of the starting BvLz was lost during the membrane processing. None of the evaluated extraction conditions caused a substantial recovery of BvLz in the concentrate. Alternative purification options for the IEX+HIC process, which achieved 95% BvLz purity, were tested. Direct loading of sugarcane extract concentrate on HIC and XAD-4 pretreatment of juice did not recovered BvLz as effectively as the IEX chromatography. Pure BvLz was obtained by the XAD+HIC process, but higher purification fold and HIC yield were achieved by the IEX+HIC process, due to the complete separation of BvLz and 18-kDa protein.

Barros, Georgia

2012-05-01T23:59:59.000Z

275

Regenerative process and system for the simultaneous removal of particulates and the oxides of sulfur and nitrogen from a gas stream  

DOE Patents (OSTI)

A process and system for simultaneously removing from a gaseous mixture, sulfur oxides by means of a solid sulfur oxide acceptor on a porous carrier, nitrogen oxides by means of ammonia gas and particulate matter by means of filtration and for the regeneration of loaded solid sulfur oxide acceptor. Finely-divided solid sulfur oxide acceptor is entrained in a gaseous mixture to deplete sulfur oxides from the gaseous mixture, the finely-divided solid sulfur oxide acceptor being dispersed on a porous carrier material having a particle size up to about 200 microns. In the process, the gaseous mixture is optionally pre-filtered to remove particulate matter and thereafter finely-divided solid sulfur oxide acceptor is injected into the gaseous The government of the United States of America has rights in this invention pursuant to Contract No. DE-AC21-88MC 23174 awarded by the U.S. Department of Energy.

Cohen, Mitchell R. (Troy, NY); Gal, Eli (Lititz, PA)

1993-01-01T23:59:59.000Z

276

Regenerative process and system for the simultaneous removal of particulates and the oxides of sulfur and nitrogen from a gas stream  

DOE Patents (OSTI)

A process and system are described for simultaneously removing from a gaseous mixture, sulfur oxides by means of a solid sulfur oxide acceptor on a porous carrier, nitrogen oxides by means of ammonia gas and particulate matter by means of filtration and for the regeneration of loaded solid sulfur oxide acceptor. Finely-divided solid sulfur oxide acceptor is entrained in a gaseous mixture to deplete sulfur oxides from the gaseous mixture, the finely-divided solid sulfur oxide acceptor being dispersed on a porous carrier material having a particle size up to about 200 microns. In the process, the gaseous mixture is optionally pre-filtered to remove particulate matter and thereafter finely-divided solid sulfur oxide acceptor is injected into the gaseous mixture.

Cohen, M.R.; Gal, E.

1993-04-13T23:59:59.000Z

277

Development and Demonstration of Waste Heat Integration with Solvent Process for More Efficient CO2 Removal from Coal-Fired Flue Gas  

NLE Websites -- All DOE Office Websites (Extended Search)

and Demonstration of and Demonstration of Waste Heat Integration with Solvent Process for More Efficient CO 2 Removal from Coal-Fired Flue Gas Background The mission of the U.S. Department of Energy/National Energy Technology Laboratory (DOE/NETL) Existing Plants, Emissions, & Capture (EPEC) Research & Development (R&D) Program is to develop innovative environmental control technologies to enable full use of the nation's vast coal reserves, while at the same time allowing the current fleet of coal-

278

The removal of uranium from acidic media using ion exchange and/or extraction chromatography  

SciTech Connect

The separation and purification of uranium from either nitric acid or hydrochloric acid media can be accomplished by using either solvent extraction or ion-exchange. Over the past two years at Los Alamos, emerging programs are focused on recapturing the expertise required to do limited, small-quantity processing of enriched uranium. During this period of time, we have been investigating ion-addition, waste stream polishing is associated with this effort in order to achieve more complete removal of uranium prior to recycle of the acid. Extraction chromatography has been demonstrated to further polish the uranium from both nitric and hydrochloric acid media thus allowing for a more complete recovery of the actinide material and creation of less waste during the processing steps.

FitzPatrick, J.R.; Schake, B.S.; Murphy, J.; Holmes, K; West, M.H.

1996-06-01T23:59:59.000Z

279

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

SciTech Connect

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 on 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.

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

2012-07-01T23:59:59.000Z

280

Overview of Fiscal Year 2002 Research and Development for Savannah River Site's Salt Waste Processing Facility  

SciTech Connect

The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste program is responsible for storage, treatment, and immobilization of high-level waste for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of the SRS high-level waste effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site's grout facility (Saltstone) and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste and sent as feed to Defense Waste Processing Facility include actinides, strontium, cesium, and entrained sludge. In fiscal year 2002 (FY02), research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility. The SPP R&D focused on advancing the technical maturity, risk reduction, engineering development, and design support for DOE's engineering, procurement, and construction (EPC) contractors for the Salt Waste Processing Facility. Thus, R&D in FY02 addressed the areas of actual waste performance, process chemistry, engineering tests of equipment, and chemical and physical properties relevant to safety. All of the testing, studies, and reports were summarized and provided to the DOE to support the Salt Waste Processing Facility, which began conceptual design in September 2002.

H. D. Harmon, R. Leugemors, PNNL; S. Fink, M. Thompson, D. Walker, WSRC; P. Suggs, W. D. Clark, Jr

2003-02-26T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Joint Actinide Shock Physics Experimental Research | National Nuclear  

National Nuclear Security Administration (NNSA)

Actinide Shock Physics Experimental Research | National Nuclear Actinide Shock Physics Experimental Research | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Jasper Joint Actinide Shock Physics Experimental Research Home > About Us > Our Programs > Defense Programs > Office of Research, Development, Test, and Evaluation > Office of Research and Development >

282

Actinide and xenon reactivity effects in ATW high flux systems  

SciTech Connect

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C. [Department of Mechanical, Aerospace, and Nuclear Engineering University of Virginia, Charlottesville, Virginia 22903 (United States); Department of Nuclear Engineering and Engineering Physics University of Wisconsin, Madison, Wisconsin 53706-1687 (United States); Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

1995-09-15T23:59:59.000Z

283

Removal of technetium from alkaline nuclear-waste media by a solvent-extraction process using crown ethers  

SciTech Connect

Crown ethers dissolved in suitably modified aliphatic kerosene diluents can be employed to extract technetium as pertechnetate anion (TcO{sub 4}{sup {minus}}) with good extraction ratios from realistic simulants of radioactive alkaline nitrate waste. The modifiers utilized are non-halogenated and non-volatile, and the technetium can be removed from the solvent by stripping using water. The crown ethers bis-4,4{prime}(5{prime})[(tert-butyl)cyclohexano]-18-crown-6 (di-t-BuCH18C6) and dicyclohexano-18-crown-6 (DCH18C6) provide stronger TcO{sub 4}{sup {minus}} extraction than dicyclohexano-21-crown-7 and 4-tert-butylcyclohexano 15-crown-5. Whereas DCH18C6 provides somewhat higher TcO{sub 4}{sup {minus}} extraction ratios than the more lipophilic di-t-BuCH18C6 derivative, the latter was selected for further study owing to its lower distribution to the aqueous phase. Particularly good extraction and stripping results were obtained with di-t-BuCH 18C6 at 0.02 M in a 2:1 vol/vol blend of tributyl phosphate and Isopar{reg_sign} M. Using this solvent, 98.9% of the technetium contained (at 6 {times} 10{sup {minus}5} M) in a Double-Shell Slurry Feed (DSSF) Hanford tank waste simulant was removed following two cross-current extraction contacts. Two cross-current stripping contacts with deionized water afforded removal of 99.1% of the technetium from the organic solvent.

Bonnesen, P.V.; Presley, D.J.; Haverlock, T.J.; Moyer, B.A. [Oak Ridge National Lab., TN (United States). Chemical and Analytical Sciences Div.

1995-07-01T23:59:59.000Z

284

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering  

DOE Patents (OSTI)

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

Vijayan, S.; Wong, C.F.; Buckley, L.P.

1994-11-22T23:59:59.000Z

285

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering  

DOE Patents (OSTI)

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved.

Vijayan, Sivaraman (Deep River, CA); Wong, Chi F. (Pembroke, CA); Buckley, Leo P. (Deep River, CA)

1994-01-01T23:59:59.000Z

286

Minor actinide waste disposal in deep geological boreholes  

E-Print Network (OSTI)

The purpose of this investigation was to evaluate a waste canister design suitable for the disposal of vitrified minor actinide waste in deep geological boreholes using conventional oil/gas/geothermal drilling technology. ...

Sizer, Calvin Gregory

2006-01-01T23:59:59.000Z

287

30th Actinide Separations Conference, PNNL-SA-50126  

SciTech Connect

Program booklet for the 30th Actinide Separations Conference. Contains agenda and abstracts for 27 poster and 38 oral presentations to be made during the 3-day meeting, May 23-25, 2006.

Delegard, Calvin H.

2006-05-25T23:59:59.000Z

288

ACTINIDE-SPECIFIC SEQUESTERING AGENTS AND DECONTAMINATION APPLICATIONS  

E-Print Network (OSTI)

In: The Health Eff, of Plutonium and Radium, Proc. Sym. ,The Metabolism of Compounds Plutonium and Other Actinides.In: The Radiobiology of Plutonium. Stover, B. J. , Jee, H.

Smith, William L.

2013-01-01T23:59:59.000Z

289

Contaminant-Organic Complexes: Their Structure and Energetics in Surface Decontamination Processes  

SciTech Connect

The current debate over possible decontamination processes for DOE facilities is centered on disparate decontamination problems, but the key contaminants (Thorium [Th],uranium [U], and plutonium [Pu]) are universally important. Innovative agents used alone or in conjunction with traditional processes can increase the potential to reclaim for future use some these valuable resources or at the least decontaminate the metal surfaces to allow disposal as nonradioactive, nonhazardous material. This debate underscores several important issues: (1) regardless of the decontamination scenario, metal (Fe, U, Pu, Np) oxide film removal from the surface is central to decontamination; and (2) simultaneous oxide dissolution and sequestration of actinide contaminants against re-adsorption to a clean metal surface will influence the efficacy of a process or agent and its cost. Current research is investigating the use of microbial siderophores (chelates) to solubilize actinides (i.e., Th, U, Pu) from the surface of Fe oxide surfaces. Continuing research integrates (1) studies of macroscopic dissolution/desorption of common actinide (IV) [Th, U, Pu, Np] solids and species sorbed to and incorporated into Fe oxides, (2) molecular spectroscopy (FTIR, Raman, XAS), to probe the structure and bonding of contaminants, siderophores and their functional moieties, and how these change with the chemical environment, (3) and molecular mechanics and electronic structure calculations to design model siderophore compounds to test and extend the MM3 model.

Ainsworth, Calvin C.; Hay, Benjamin P.; Traina, Samuel J.; Myneni, Satish C. B.

2003-06-01T23:59:59.000Z

290

Use of the TRUEX process for the pretreatment of neutralized cladding removal waste (NCRW) sludge: Results of a design basis experiment  

SciTech Connect

This report presents the results of an experiment designed to demonstrate the feasibility of a sludge dissolution/solvent extraction process to separate transuranic elements from the bulk components of Hanford neutralized cladding removal waste (NCRW) sludge. Such a separation would allow the bulk of the waste to be disposed of as low-level waste, which is much less costly than geologic disposal as would be required for the waste in its current form. The results indicate that the proposed process is well suited to meet the desired objectives. A composite sample of NCRW sludge taken from Tank 103-AW in 1986 was dissolved in nitric acid at room temperature. Dissolution of bulk components and all radionuclides was {ge}95% complete; thus, {le}5% of the bulk components will require geologic disposal. The TRUEX (TRansUranium EXtraction) solvent extraction process gave very good separation of the transuranic from the bulk components of the waste.

Swanson, J L

1991-07-01T23:59:59.000Z

291

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering  

DOE Patents (OSTI)

It is an object of the claimed invention to combine chemical treatment with microfiltration process to treat groundwater, leachate from contaminated soil washing, surface and run-off waters contaminated with toxic metals, radionuclides and trace amounts of organics from variety of sources. The process can also be used to treat effluents from industrial processes such as discharges associated with smelting, mining and refining operations. Influent contaminants amenable to treatment are from a few mg/L to hundreds of mg/L. By selecting appropriate precipitation, ion exchange and adsorption agents and conditions, efficiencies greater than 99.9 percent can be achieved for removal of contaminants. The filtered water for discharge can be targeted with either an order of magnitude greater or lower than contaminant levels for drinking water.

Vijayan, S.; Wong, Chi Fun; Buckley, L.P.

1992-12-31T23:59:59.000Z

292

Ion Removal  

INL’s ion removal technology leverages the ability of phosphazene polymers discriminate between water and metal ions, which allows water to pass ...

293

Gas separation process using membranes with permeate sweep to remove CO.sub.2 from gaseous fuel combustion exhaust  

DOE Patents (OSTI)

A gas separation process for treating exhaust gases from the combustion of gaseous fuels, and gaseous fuel combustion processes including such gas separation. The invention involves routing a first portion of the exhaust stream to a carbon dioxide capture step, while simultaneously flowing a second portion of the exhaust gas stream across the feed side of a membrane, flowing a sweep gas stream, usually air, across the permeate side, then passing the permeate/sweep gas back to the combustor.

Wijmans Johannes G. (Menlo Park, CA); Merkel, Timothy C. (Menlo Park, CA); Baker, Richard W. (Palo Alto, CA)

2012-05-15T23:59:59.000Z

294

Actinide Production in the Reaction of Heavy Ions withCurium-248  

Science Conference Proceedings (OSTI)

Chemical experiments were performed to examine the usefulness of heavy ion transfer reactions in producing new, neutron-rich actinide nuclides. A general quasi-elastic to deep-inelastic mechanism is proposed, and the utility of this method as opposed to other methods (e.g. complete fusion) is discussed. The relative merits of various techniques of actinide target synthesis are discussed. A description is given of a target system designed to remove the large amounts of heat generated by the passage of a heavy ion beam through matter, thereby maximizing the beam intensity which can be safely used in an experiment. Also described is a general separation scheme for the actinide elements from protactinium (Z = 91) to mendelevium (Z = 101), and fast specific procedures for plutonium, americium and berkelium. The cross sections for the production of several nuclides from the bombardment of {sup 248}Cm with {sup 18}O, {sup 86}Kr and {sup 136}Xe projectiles at several energies near and below the Coulomb barrier were determined. The results are compared with yields from {sup 48}Ca and {sup 238}U bombardments of {sup 248}Cm. Simple extrapolation of the product yields into unknown regions of charge and mass indicates that the use of heavy ion transfer reactions to produce new, neutron-rich above-target species is limited. The substantial production of neutron-rich below-target species, however, indicates that with very heavy ions like {sup 136}Xe and {sup 238}U the new species {sup 248}Am, {sup 249}Am and {sup 247}Pu should be produced with large cross sections from a {sup 248}Cm target. A preliminary, unsuccessful attempt to isolate {sup 247}Pu is outlined. The failure is probably due to the half life of the decay, which is calculated to be less than 3 minutes. The absolute gamma ray intensities from {sup 251}Bk decay, necessary for calculating the {sup 251}Bk cross section, are also determined.

Moody, K.J.

1983-07-01T23:59:59.000Z

295

Actinide Sorption in Rainier Mesa Tunnel Waters from the Nevada Test Site  

SciTech Connect

The sorption behavior of americium (Am), plutonium (Pu), neptunium (Np), and uranium (U) in perched Rainier Mesa tunnel water was investigated. Both volcanic zeolitized tuff samples and groundwater samples were collected from Rainier Mesa, Nevada Test Site, NV for a series of batch sorption experiments. Sorption in groundwater with and without the presence of dissolved organic matter (DOM) was investigated. Am(III) and Pu(IV) are more soluble in groundwater that has high concentrations of DOM. The sorption K{sub d} for Am(III) and Pu(IV) on volcanic zeolitized tuff was up to two orders of magnitude lower in samples with high DOM (15 to 19 mg C/L) compared to samples with DOM removed (< 0.4 mg C/L) or samples with naturally low DOM (0.2 mg C/L). In contrast, Np(V) and U(VI) sorption to zeolitized tuff was much less affected by the presence of DOM. The Np(V) and U(VI) sorption Kds were low under all conditions. Importantly, the DOM was not found to significantly sorb to the zeolitized tuff during these experiment. The concentration of DOM in groundwater affects the transport behavior of actinides in the subsurface. The mobility of Am(III) and Pu(IV) is significantly higher in groundwater with elevated levels of DOM resulting in potentially enhanced transport. To accurately model the transport behavior of actinides in groundwater at Rainier Mesa, the low actinide Kd values measured in groundwater with high DOM concentrations must be incorporated in predictive transport models.

Zhao, P; Zavarin, M; Leif, R; Powell, B; Singleton, M; Lindvall, R; Kersting, A

2007-12-17T23:59:59.000Z

296

Sigma Team for Minor Actinide Separation: PNNL FY 2010 Status Report  

SciTech Connect

Work conducted at Pacific Northwest National Laboratory (PNNL) in FY 2010 addressed two lines of inquiry. The two hypotheses put forth were: 1. The extractants from the TRUEX( ) process (CMPO)( ) and from the TALSPEAK( ) process (HDEHP)( ) can be combined into a single process solvent to separate 1) the lanthanides and actinides from acidic high-level waste and 2) the actinides from the lanthanides in a single solvent extraction process. (Note: This combined process will hereafter be referred to as the TRUSPEAK process.) A series of empirical measurements performed (both at PNNL and Argonne National Laboratory) in FY 2009 supported this hypothesis, but also indicated some nuances to the chemistry. Lanthanide/americium separation factors of 12 and higher were obtained with a prototypic TRUSPEAK solvent when extracting the lanthanides from a citrate-buffered DTPA( ) solution. Although the observed separation factors are sufficiently high to design an actinide/lanthanide separation process, a better understanding of the chemistry is expected to lead to improved solvent formulations and improved process performance. Work in FY 2010 focused on understanding the synergistic extraction behavior observed for Nd(III) and Am(III) when extracted into mixtures of CMPO and HDEHP. The interaction between CMPO and HDEHP in dodecane was investigated by 31P NMR spectroscopy, and an adduct of the type CMPO•HDEHP was found to form. The formation of this adduct will reduce the effective extractant concentrations and must be taken into account when modeling metal ion extraction data in this system. Studies were also initiated to determine the Pitzer parameters for Nd(III) in lactate media. 2. Higher oxidation states (e.g., +5 and +6) of Am can be stabilized in solution by complexation with uranophilic ligands, and this chemistry can be exploited to separate Am from Cm. To test this hypothesis, the previously reported stereognostic uranophilic ligands NPB( ) and ETAC(e) were investigated. To assess the potential of these ligands to stabilize pentavalent and hexavalent actinides, stability constants were measured for complexation of these ligands to Nd(III), Np(V), and Pu(VI) in a solvent mixture consisting of 80% methanol/20% water. For comparison, an analogous non-stereognostic ligand, NTA,(f) was also examined. The ligand ETAC showed greater binding affinity for hexavalent Pu versus trivalent Nd by two orders-of-magnitude. Such selectivity was not observed for either NTA or NPB.

Lumetta, Gregg J.; Sinkov, Sergey I.; Neiner, Doinita; Levitskaia, Tatiana G.; Braley, Jenifer C.; Carter, Jennifer C.; Warner, Marvin G.; Pittman, Jonathan W.; Rapko, Brian M.

2010-08-24T23:59:59.000Z

297

Dynamic Recovery in Silicate-Apatite Structures Under Irradiation and Implications for Long-Term Immobilization of Actinides  

SciTech Connect

The irradiation responses of Ca{sub 2}La{sub 8}(SiO{sub 4}){sub 6}O{sub 2} and Sr{sub 2}Nd{sub 8}(SiO{sub 4}){sub 6}O{sub 2} with the apatite structure are investigated to predict their long-term behaviour as host phases for immobilization of actinide elements from the nuclear fuel cycle. Different ions and energies are used to study the effects of dose, temperature, atomic displacement rate and ionization rate on irradiation-induced amorphization and recrystallization. The dose for amorphization increases with temperature in two stages, below and above 150 K. In the high temperature stage relevant to actinide immobilization, the increase of amorphization dose with temperature exhibits a strong dependence on the ratio of ionization rate to displacement rate for the different ions. Data analysis using a dynamic model for amorphization reveals that ionization-induced processes, with activation energy of 0.15 {+-} 0.02 eV, dominate dynamic recovery for ions from Ne through Xe. For heavier Au ions or for alpha-recoil nuclei emitted in alpha decay of actinides, ionization becomes less dominant and dynamic recovery is controlled primarily by thermally-driven processes. In post-irradiation annealing studies of amorphous samples, epitaxial thermal recrystallization is observed at 1123 K, and irradiation-enhanced nucleation of nanocrystallites is observed under irradiation with heavier ions. The recrystallization temperature under irradiation decreases with increasing ion mass to a value of {approx} 823 K, which also defines the thermally-driven critical temperature for amorphization under irradiation with heavy ions. Some partial recovery due to alpha particle irradiation at 300 K is observed that suggests a self-healing mechanism in apatite phases containing actinides. Based on the results and dynamic model, the temperature and time dependences of amorphization in silicate-apatite host phases for actinide immobilization are predicted.

Weber, William J.; Zhang, Yanwen; Xiao, Haiyan Y.; Wang, Lumin M.

2011-11-14T23:59:59.000Z

298

Dynamic recovery in silicate-apatite structures under irradiation and implications for long-term immobilization of actinides  

SciTech Connect

The irradiation responses of Ca2La8(SiO4)6O2 and Sr2Nd8(SiO4)6O2 with the apatite structure are investigated to predict their long-term behaviour as host phases for immobilization of actinide elements from the nuclear fuel cycle. Different ions and energies are used to study the effects of dose, temperature, atomic displacement rate and ionization rate on irradiation-induced amorphization and recrystallization. The dose for amorphization increases with temperature in two stages, below and above 150 K. In the high temperature stage relevant to actinide immobilization, the increase of amorphization dose with temperature exhibits a strong dependence on the ratio of ionization rate to displacement rate for the different ions. Data analysis using a dynamic model for amorphization reveals that ionization-induced processes, with activation energy of 0.15 0.02 eV, dominate dynamic recovery for ions from Ne through Xe. For heavier Au ions or for alpha-recoil nuclei emitted in alpha decay of actinides, ionization becomes less dominant and dynamic recovery is controlled primarily by thermally-driven processes. In post-irradiation annealing studies of amorphous samples, epitaxial thermal recrystallization is observed at 1123 K, and irradiation-enhanced nucleation of nanocrystallites is observed under irradiation with heavier ions. The recrystallization temperature under irradiation decreases with increasing ion mass to a value of ~ 823 K, which also defines the thermally-driven critical temperature for amorphization under irradiation with heavy ions. Some partial recovery due to alpha particle irradiation at 300 K is observed that suggests a self-healing mechanism in apatite phases containing actinides. Based on the results and dynamic model, the temperature and time dependences of amorphization in silicate-apatite host phases for actinide immobilization are predicted.

Weber, William J [ORNL; Zhang, Yanwen [ORNL; Xiao, Haiyan [University of Tennessee, Knoxville (UTK); Wang, Prof. Lumin [University of Michigan

2012-01-01T23:59:59.000Z

299

A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel  

Science Conference Proceedings (OSTI)

An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

Samuel E. Bays

2011-02-01T23:59:59.000Z

300

CRITICALITY SAFETY OF PROCESSING SALT SOLUTION AT SRS  

Science Conference Proceedings (OSTI)

High level radioactive liquid waste generated as a result of the production of nuclear material for the United States defense program at the Savannah River Site has been stored as 36 million gallons in underground tanks. About ten percent of the waste volume is sludge, composed of insoluble metal hydroxides primarily hydroxides of Mn, Fe, Al, Hg, and most radionuclides including fission products. The remaining ninety percent of the waste volume is saltcake, composed of primarily sodium (nitrites, nitrates, and aluminates) and hydroxides. Saltcakes account for 30% of the radioactivity while the sludge accounts for 70% of the radioactivity. A pilot plant salt disposition processing system has been designed at the Savannah River Site for interim processing of salt solution and is composed of two facilities: the Actinide Removal Process Facility (ARPF) and the Modular Caustic Side Solvent Extraction Unit (MCU). Data from the pilot plant salt processing system will be used for future processing salt at a much higher rate in a new salt processing facility. Saltcake contains significant amounts of actinides, and other long-lived radioactive nuclides such as strontium and cesium that must be extracted prior to disposal as low level waste. The extracted radioactive nuclides will be mixed with the sludge from waste tanks and vitrified in another facility. Because of the presence of highly enriched uranium in the saltcake, there is a criticality concern associated with concentration and/or accumulation of fissionable material in the ARP and MCU.

Stephens, K; Davoud Eghbali, D; Michelle Abney, M

2008-01-15T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Method and apparatus for removing and preventing window deposition during photochemical vapor deposition (photo-CVD) processes  

DOE Patents (OSTI)

Unwanted build-up of the film deposited on the transparent light-transmitting window of a photochemical vacuum deposition (photo-CVD) chamber is eliminated by flowing an etchant into the part of the photolysis region in the chamber immediately adjacent the window and remote from the substrate and from the process gas inlet. The respective flows of the etchant and the process gas are balanced to confine the etchant reaction to the part of the photolysis region proximate to the window and remote from the substrate. The etchant is preferably one that etches film deposit on the window, does not etch or affect the window itself, and does not produce reaction by-products that are deleterious to either the desired film deposited on the substrate or to the photolysis reaction adjacent the substrate.

Tsuo, Simon (Lakewood, CO); Langford, Alison A. (Boulder, CO)

1989-01-01T23:59:59.000Z

302

Method and apparatus for removing and preventing window deposition during photochemical vapor deposition (photo-CVD) processes  

DOE Patents (OSTI)

Unwanted build-up of the film deposited on the transparent light-transmitting window of a photochemical vacuum deposition (photo-CVD) chamber is eliminated by flowing an etchant into the part of the photolysis region in the chamber immediately adjacent the window and remote from the substrate and from the process gas inlet. The respective flows of the etchant and the process gas are balanced to confine the etchant reaction to the part of the photolysis region proximate to the window and remote from the substrate. The etchant is preferably one that etches film deposit on the window, does not etch or affect the window itself, and does not produce reaction by-products that are deleterious to either the desired film deposited on the substrate or to the photolysis reaction adjacent the substrate. 3 figs.

Tsuo, S.; Langford, A.A.

1989-03-28T23:59:59.000Z

303

DEVELOPMENT PROGRAM FOR PU-238 AQUEOUS RECOVERY PROCESS  

SciTech Connect

Aqueous processing is necessary for the removal of impurities from {sup 238}Pu dioxide ({sup 238}PuO{sub 2}) fuel due to unacceptable levels of {sup 234}U and other non-actinide impurities in the scrap fuel. Impurities at levels above General Purpose Heat Source (GPHS) fuel specifications may impair the performance.of the heat sources. Efforts at Los Alamos have focused on developing the bench scale methodology for the aqueous process steps which includes comminution, dissolution, ion exchange, precipitation, and calcination. Recently, work has been performed to qualify the bench scale methodology, to show that the developed process produces pure {sup 238}PuO{sub 2} meeting GPHS fuel specifications. In addition, this work has enabled us to determine how waste volumes may be minimized during full-scale processing. Results of process qualification for the bench scale aqueous recovery operation and waste minimization efforts are presented.

M. PANSOY-HJELVIK; M. REIMUS; ET AL

2000-10-01T23:59:59.000Z

304

The TRansUranium EXtraction (TRUEX) process: A vital tool for disposal of US defense nuclear waste  

SciTech Connect

The TRUEX (TRansUranium EXtraction) process is a generic actinide extraction/recovery process for the removal of all actinides from acidic nitrate and chloride nuclear waste solutions. Because of its high efficiency and flexibility and its compatibility with existing process facilities, TRUEX has now become a vital tool for the disposal of certain US defense nuclear waste. The development of TRUEX is closely coupled to the development of bifunctional extractants belonging to the carbamoylphosphoryl class and CMPO in particular. A brief review of the development of CMPO and its relationship to other bifunctional and monofunctional extractants is presented. The effect of TBP on CMPO, the selectivity of CMPO for actinides extracted from acidic nitrate media, the influence of diluents on CMPO behavior and 3rd phase formation, and the radiolysis/hydrolysis of CMPO and subsequent solvent cleanup will be highlighted. Application of TRUEX in the chemical pretreatment of specific nuclear waste streams and a summary of the current status of development and deployment of TRUEX is presented. 15 refs., 10 figs., 3 tabs.

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

1990-01-01T23:59:59.000Z

305

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies  

SciTech Connect

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.

Kenneth Allen; Travis Knight; Samuel Bays

2011-05-01T23:59:59.000Z

306

Use of once-through treat gas to remove the heat of reaction in solvent hydrogenation processes  

DOE Patents (OSTI)

In a coal liquefaction process wherein feed coal is contacted with molecular hydrogen and a hydrogen-donor solvent in a liquefaction zone to form coal liquids and vapors and coal liquids in the solvent boiling range are thereafter hydrogenated to produce recycle solvent and liquid products, the improvement which comprises separating the effluent from the liquefaction zone into a hot vapor stream and a liquid stream; cooling the entire hot vapor stream sufficiently to condense vaporized liquid hydrocarbons; separating condensed liquid hydrocarbons from the cooled vapor; fractionating the liquid stream to produce coal liquids in the solvent boiling range; dividing the cooled vapor into at least two streams; passing the cooling vapors from one of the streams, the coal liquids in the solvent boiling range, and makeup hydrogen to a solvent hydrogenation zone, catalytically hydrogenating the coal liquids in the solvent boiling range and quenching the hydrogenation zone with cooled vapors from the other cooled vapor stream.

Nizamoff, Alan J. (Convent Station, NJ)

1980-01-01T23:59:59.000Z

307

Theoretical Design of a Thermosyphon for Efficient Process Heat Removal from Next Generation Nuclear Plant (NGNP) for Production of Hydrogen  

DOE Green Energy (OSTI)

The work reported here is the preliminary analysis of two-phase Thermosyphon heat transfer performance with various alkali metals. Thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. Heat transport occurs via evaporation and condensation, and the heat transport fluid is re-circulated by gravitational force. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. For process heat, intermediate heat exchangers (IHX) are required to transfer heat from the NGNP to the hydrogen plant in the most efficient way possible. The production of power at higher efficiency using Brayton Cycle, and hydrogen production requires both heat at higher temperatures (up to 1000oC) and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. The purpose for selecting a compact heat exchanger is to maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. The IHX design requirements are governed by the allowable temperature drop between the outlet of the NGNP (900oC, based on the current capabilities of NGNP), and the temperatures in the hydrogen production plant. Spiral Heat Exchangers (SHE’s) have superior heat transfer characteristics, and are less susceptible to fouling. Further, heat losses to surroundings are minimized because of its compact configuration. SHEs have never been examined for phase-change heat transfer applications. The research presented provides useful information for thermosyphon design and Spiral Heat Exchanger.

Piyush Sabharwall; Fred Gunnerson; Akira Tokuhiro; Vivek Utgiker; Kevan Weaver; Steven Sherman

2007-10-01T23:59:59.000Z

308

Sigma Team for Minor Actinide Separation: PNNL FY 2011 Status Report  

SciTech Connect

This report summarizes work conducted in FY 2011 at PNNL to investigate new methods of separating the minor actinide elements (Am and Cm) from the trivalent lanthanide elements, and separation of Am from Cm. For the former, work focused on a solvent extraction system combining an acidic extractant (HDEHP) with a neutral extractant (CMPO) to form a hybrid solvent extraction system referred to as TRUSPEAK (combining the TRUEX and TALSPEAK processes). For the latter, ligands that strongly bing uranyl ion were investigated for stabilizing corresponding americyl ion.

Lumetta, Gregg J.; Braley, Jenifer C.; Sinkov, Sergey I.; Levitskaia, Tatiana G.; Carter, Jennifer C.; Warner, Marvin G.; Pittman, Jonathan W.

2011-08-13T23:59:59.000Z

309

Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

Timothy A. Hyde

2012-06-01T23:59:59.000Z

310

Environmental Assessment for Actinide Chemistry and Repository Science  

NLE Websites -- All DOE Office Websites (Extended Search)

questions on the Environmental Assessment for Actinide Chemistry and Repository Science Laboratory, email Harold.Johnson@wipp.ws or call (505) 234-7349. questions on the Environmental Assessment for Actinide Chemistry and Repository Science Laboratory, email Harold.Johnson@wipp.ws or call (505) 234-7349. Environmental Assessment for Actinide Chemistry and Repository Science Laboratory Final - January, 2006 This document has been provided to you in PDF format. Please install Adobe Acrobat Reader before accessing these documents. Some of the Chapters containing complex graphics have been split into multiple parts to allow for more detail in the graphics and ease in downloading. Cover Sheet, Table of Contents, List of Tables, List of Figures, and Acronyms and Abbreviations Chapter 1 - Introduction and Statement of Purpose and Need Chapter 2 - Proposed Action and Alternatives Chapter 3 - Existing Environment

311

FY2010 Annual Report for the Actinide Isomer Detection Project  

Science Conference Proceedings (OSTI)

This project seeks to identify a new signature for actinide element detection in active interrogation. This technique works by exciting and identifying long-lived nuclear excited states (isomers) in the actinide isotopes and/or primary fission products. Observation of isomers in the fission products will provide a signature for fissile material. For the actinide isomers, the decay time and energy of the isomeric state is unique to a particular isotope, providing an unambiguous signature for Special Nuclear Materials (SNM). Future work will include a follow-up measurement scheduled for December 2010 at LBNL. Lessons learned from the July 2010 measurements will be incorporated into these new measurements. Analysis of both the July and December experiments will be completed in a few months. A research paper to be submitted to a peer-reviewed journal will be drafted if the conclusions from the measurements warrant publication.

Warren, Glen A.; Francy, Christopher J.; Ressler, Jennifer J.; Erikson, Luke E.; Miller, Erin A.; Hatarik, R.

2011-01-01T23:59:59.000Z

312

Actinide (III) solubility in WIPP Brine: data summary and recommendations  

Science Conference Proceedings (OSTI)

The solubility of actinides in the +3 oxidation state is an important input into the Waste Isolation Pilot Plant (WIPP) performance assessment (PA) models that calculate potential actinide release from the WIPP repository. In this context, the solubility of neodymium(III) was determined as a function of pH, carbonate concentration, and WIPP brine composition. Additionally, we conducted a literature review on the solubility of +3 actinides under WIPP-related conditions. Neodymium(III) was used as a redox-invariant analog for the +3 oxidation state of americium and plutonium, which is the oxidation state that accounts for over 90% of the potential release from the WIPP through the dissolved brine release (DBR) mechanism, based on current WIPP performance assessment assumptions. These solubility data extend past studies to brine compositions that are more WIPP-relevant and cover a broader range of experimental conditions than past studies.

Borkowski, Marian; Lucchini, Jean-Francois; Richmann, Michael K.; Reed, Donald T.

2009-09-01T23:59:59.000Z

313

Actinide Sorption in Rainier Mesa Tunnel Waters from the Nevada Test Site  

Science Conference Proceedings (OSTI)

The sorption behavior of americium (Am), plutonium (Pu), neptunium (Np), and uranium (U) in perched Rainier Mesa tunnel water was investigated. Both volcanic zeolitized tuff samples and groundwater samples were collected from Rainier Mesa, Nevada Test Site, NV for a series of batch sorption experiments. Sorption in groundwater with and without the presence of dissolved organic matter (DOM) was investigated. Am(III) and Pu(IV) are more soluble in groundwater that has high concentrations of DOM. The sorption K{sub d} for Am(III) and Pu(IV) on volcanic zeolitized tuff was up to two orders of magnitude lower in samples with high DOM (15 to 19 mg C/L) compared to samples with DOM removed (Rainier Mesa, the low actinide Kd values measured in groundwater with high DOM concentrations must be incorporated in predictive transport models.

Zhao, P; Zavarin, M; Leif, R; Powell, B; Singleton, M; Lindvall, R; Kersting, A

2007-12-17T23:59:59.000Z

314

Selection of actinide chemical analogues for WIPP tests  

Science Conference Proceedings (OSTI)

The Department of Energy must demonstrate the effectiveness of the Waste Isolation Pilot Plant (WIPP) as a permanent repository for the disposal of transuranic (TRU) waste. Performance assessments of the WIPP require that estimates of the transportability and outcome of the radionuclides (actinides) be determined from disposal rooms that may become either partially or completely filled with brine. Federal regulations limit the amount of radioactivity that may be unintentionally released to the accessible environment by any mechanism during the post closure phase up to 10,000 years. Thermodynamic models have been developed to predict the concentrations of actinides in the WIPP disposal rooms under various situations and chemical conditions. These models are based on empirical and theoretical projections of the chemistry that might be present in and around the disposal room zone for both near and long-term periods. The actinides that are known to be present in the TRU wastes (and are included in the model) are Th, U, Np, Pu, and Am. Knowledge of the chemistry that might occur in the disposal rooms when the waste comes in contact with brine is important in understanding the range of oxidation states that might be present under different conditions. There is a need to establish the mechanisms and resultant rate of transport, migration, or effective retardation of actinides beyond the disposal rooms to the boundary of the accessible environment. The influence of the bulk salt rock, clay sediments and other geologic matrices on the transport behavior of actinides must be determined to establish the overall performance and capability of the WIPP in isolating waste from the environment. Tests to determine the capabilities of the WIPP geologic formations in retarding actinide species in several projected oxidation states would provide a means to demonstrate the effectiveness of the WIPP in retaining TRU wastes.

Villarreal, R.; Spall, D.

1995-07-05T23:59:59.000Z

315

Physics studies of higher actinide consumption in an LMR  

Science Conference Proceedings (OSTI)

The core physics aspects of the transuranic burning potential of the Integral Fast Reactor (IFR) are assessed. The actinide behavior in fissile self-sufficient IFR closed cycles of 1200 MWt size is characterized, and the transuranic isotopics and risk potential of the working inventory are compared to those from a once-through LWR. The core neutronic performance effects of rare-earth impurities present in the recycled fuel are addressed. Fuel cycle strategies for burning transuranics from an external source are discussed, and specialized actinide burner designs are described. 4 refs., 4 figs., 3 tabs.

Hill, R.N.; Wade, D.C.; Fujita, E.K.; Khalil, H.S.

1990-01-01T23:59:59.000Z

316

Demonstration of EIC's copper sulfate process for removal of hydrogen sulfide and other trace contaminants from geothermal steam at turbine inlet temperatures and pressures. Final report  

DOE Green Energy (OSTI)

The results obtained during the operation of an integrated, one-tenth commercial scale pilot plant using EIC's copper sulfate process for the removal of hydrogen sulfide and other contaminants from geothermal steam at turbine upstream conditions are discussed. The tests took place over a six month period at Pacific Gas and Electric Company's Unit No. 7 at The Geysers Power Plant. These tests were the final phase of a development effort which included the laboratory research and engineering design work which led to the design of the pilot plant. Broadly, the objectives of operating the pilot plant were to confirm the preliminary design criteria which had been developed, and provide data for their revisions, if appropriate, in a plant which contained all the elements of a commercial process using equipment of a size sufficient to provide valid scale-up data. The test campaign was carried out in four phases: water testing; open circuit, i.e., non integrated scrubbing, liquid-solid separation and regeneration testing; closed circuit short term; and closed circuit long term testing.

Not Available

1980-05-01T23:59:59.000Z

317

Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data  

Science Conference Proceedings (OSTI)

Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

NONE

1997-11-01T23:59:59.000Z

318

Summary - SRS Salt Waste Processing Facility  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SRS Co SRS Co DOE S Proces concen actinid in a se remov adjustm sorben sorben solutio passed separa stream extract sufficie separa (with S vitrifica (DWP Sr/acti federa assure and ha Critica The te (CTE) descrip Readin The Ele Site: S roject: S F Report Date: J ited States Why DOE omposite High Lev Savannah Rive ssing Facility (S ntrate targeted des) from High eries of unit ope ved by contactin ment) with a m nt in a batch m nt (containing S on by cross flow d to a solvent e ated to an aque m. The bulk so tion process, w ently low levels ated high activi Sr and actinide ation in the Def F). Provisions inides adsorpti al project direct e that the plann ave been matu al Decision-3 ap What th eam identified e of the SWPF w ption. All CTE ness Level of 6 To view the full T http://www.em.doe. objective of a Tech ements (CTEs), usin

319

Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels  

Science Conference Proceedings (OSTI)

The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: ? Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels. ? Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

Morgan, Dane; Yang, Yong

2013-10-28T23:59:59.000Z

320

Method for the recovery of actinide elements from nuclear reactor waste  

DOE Green Energy (OSTI)

A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid.

Horwitz, E. Philip (Elmhurst, IL); Delphin, Walter H. (Woodridge, IL); Mason, George W. (Clarendon Hills, IL)

1979-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Argonne Chemical Sciences & Engineering - Facilities - Actinide...  

NLE Websites -- All DOE Office Websites (Extended Search)

Fundamental Interactions Catalysis & Energy Conversion Electrochemical Energy Storage Nuclear & Environmental Processes National Security Institute for Atom-Efficient Chemical...

322

Special Review of the Rocky Flats Closure Project Site  

NLE Websites -- All DOE Office Websites (Extended Search)

1999, involve removing and processing actinide liquids; removing and packaging facility sludge; removing and size- reducing all tanks, piping, gloveboxes, and associated...

323

Chemical properties of the heavier actinides and transactinides  

Science Conference Proceedings (OSTI)

The chemical properties of each of the elements 99 (Es) through 105 are reviewed and their properties correlated with the electronic structure expected for 5f and 6d elements. A major feature of the heavier actinides, which differentiates them from the comparable lanthanides, is the increasing stability of the divalent oxidation state with increasing atomic number. The divalent oxidation state first becomes observable in the anhydrous halides of californium and increases in stability through the series to nobelium, where this valency becomes predominant in aqueous solution. In comparison with the analogous 4f electrons, the 5f electrons in the latter part of the series are more tightly bound. Thus, there is a lowering of the 5f energy levels with respect to the Fermi level as the atomic number increases. The metallic state of the heavier actinides has not been investigated except from the viewpoint of the relative volatility among members of the series. In aqueous solutions, ions of these elements behave as a normal trivalent actinides and lanthanides (except for nobelium). Their ionic radii decrease with increasing nuclear charge which is moderated because of increased screening of the outer 6p electrons by the 5f electrons. The actinide series of elements is completed with the element lawrencium (Lr) in which the electronic configuration is 5f/sup 14/7s/sup 2/7p. From Mendeleev's periodicity and Dirac-Fock calculations, the next group of elements is expected to be a d-transition series corresponding to the elements Hf through Hg. The chemical properties of elements 104 and 105 only have been studied and they indeed appear to show the properties expected of eka-Hf and eka-Ta. However, their nuclear lifetimes are so short and so few atoms can be produced that a rich variety of chemical information is probably unobtainable.

Hulet, E.K.

1981-01-01T23:59:59.000Z

324

Impact of actinide recycle on nuclear fuel cycle health risks  

SciTech Connect

The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR){sup 1} and Integral Fast Reactor (IF){sup 2} technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle.

Michaels, G.E.

1992-06-01T23:59:59.000Z

325

Benefits of actinide-only burnup credit for shutdown PWRs  

Science Conference Proceedings (OSTI)

Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit, analysis of the number of canisters required with and without burnup credit was made. In order to perform the analysis, loading curves were developed for the Holtec Hi-Star 100/MPC-32. The MPC-32 is hoped to be representative of future burnup credit designs from many vendors. The loading curves were generated using the actinide-only burnup credit currently under NRC review. The canister was analyzed for full loading (32 assemblies) and with partial loadings of 30 and 28 assemblies. If no burnup credit is used the maximum capacity was assumed to be 24 assemblies. this reduced capacity is due to the space required for flux traps which are needed to sufficiently reduce the canister reactivity for the fresh fuel assumption. Without burnup credit the 1,343 assemblies would require 60 canisters. If all the fuel could be loaded into the 32 assembly canisters only 45 canisters would be required. Although the actinide-only burnup credit approach is very conservative, the total number of canisters required is only 47 which is only two short of the minimum possible number of canisters. The utility is expected to buy the canister and the storage overpack. A reasonable cost estimate for the canister plus overpack is $500,000. Actinide-only burnup credit would save 13 canisters and overpacks which is a savings of about $6.5 million. This savings is somewhat reduced since burnup credit requires a verification measurement of burnup. The measurement costs for these assemblies can be estimated as about $1 million. The net savings would be $5.5 million.

Lancaster, D.; Fuentes, E.; Kang, C. [TRW Environmental Safety Systems, INc., Las Vegas, NV (United States); Rivard, D. [Maine Yankee Atomic Power Co., Westboro, MA (United States)

1998-02-01T23:59:59.000Z

326

Determination of actinides in urine and fecal samples  

DOE Patents (OSTI)

A method of determining the radioactivity of specific actinides that are carried in urine or fecal sample material is disclosed. The samples are ashed in a muffle furnace, dissolved in an acid, and then treated in a series of steps of reduction, oxidation, dissolution, and precipitation, including a unique step of passing a solution through a chloride form anion exchange resin for separation of uranium and plutonium from americium.

McKibbin, T.T.

1992-12-31T23:59:59.000Z

327

Chemical and Ceramic Methods Toward Safe Storage of Actinides  

Science Conference Proceedings (OSTI)

A very import, extremely-long-term, use for monazite as a radwaste encapsulant has been proposed. THe use of ceramic La-monazite for sequestering actinides (isolating them from the environment), especially plutonium and some other radioactive elements )e.g., fission-product rare earths), had been especially championed by Lynn Boatner of ORNL. Monazite may be used alone or, copying its compatibility with many other minerals in nature, may be used in diverse composite combinations.

P.E.D. Morgan; R.M. Housley; J.B. Davis; M.L. DeHaan

2005-08-19T23:59:59.000Z

328

Detection of Actinides via Nuclear Isomer De-Excitation  

Science Conference Proceedings (OSTI)

This dissertation discusses a data collection experiment within the Actinide Isomer Identification project (AID). The AID project is the investigation of an active interrogation technique that utilizes nuclear isomer production, with the goal of assisting in the interdiction of illicit nuclear materials. In an attempt to find and characterize isomers belonging to 235U and its fission fragments, a 232Th target was bombarded with a monoenergetic 6Li ion beam, operating at 45 MeV.

Francy, Christopher J.

2009-07-22T23:59:59.000Z

329

Determination of actinides in urine and fecal samples  

DOE Patents (OSTI)

A method of determining the radioactivity of specific actinides that are carried in urine or fecal sample material is disclosed. The samples are ashed in a muffle furnace, dissolved in an acid, and then treated in a series of steps of reduction, oxidation, dissolution, and precipitation, including a unique step of passing a solution through a chloride form anion exchange resin for separation of uranium and plutonium from americium.

McKibbin, T.T.

1993-03-02T23:59:59.000Z

330

Method for the concentration and separation of actinides from biological and environmental samples  

DOE Patents (OSTI)

A method and apparatus for the quantitative recover of actinide values from biological and environmental sample by passing appropriately prepared samples in a mineral acid solution through a separation column of a dialkyl(phenyl)-N,N-dialylcarbamoylmethylphosphine oxide dissolved in tri-n-butyl phosphate on an inert substrate which selectively extracts the actinide values. The actinide values can be eluted either as a group or individually and their presence quantitatively detected by alpha counting. 3 figs.

Horwitz, E.P.; Dietz, M.L.

1989-05-30T23:59:59.000Z

331

Method for the concentration and separation of actinides from biological and environmental samples  

DOE Patents (OSTI)

A method and apparatus for the quantitative recover of actinide values from biological and environmental sample by passing appropriately prepared samples in a mineral acid solution through a separation column of a dialkyl(phenyl)-N,N-dialylcarbamoylmethylphosphine oxide dissolved in tri-n-butyl phosphate on an inert substrate which selectively extracts the actinide values. The actinide values can be eluted either as a group or individually and their presence quantitatively detected by alpha counting.

Horwitz, E. Philip (Naperville, IL); Dietz, Mark L. (Tucson, AZ)

1989-01-01T23:59:59.000Z

332

Biomimetic Actinide Chelators: An Update on the Preclinical Development of the Orally Active Hydroxypyridonate Decorporation Agents 3,4,3-LI(1,2-HOPO) and 5-LIO(Me-3,2-HOPO)  

Science Conference Proceedings (OSTI)

The threat of a dirty bomb or other major radiological contamination presents a danger of large-scale radiation exposure of the population. Because major components of such contamination are likely to be actinides, actinide decorporation treatments that will reduce radiation exposure must be a priority. Current therapies for the treatment of radionuclide contamination are limited and extensive efforts must be dedicated to the development of therapeutic, orally bioavailable, actinide chelators for emergency medical use. Using a biomimetic approach based on the similar biochemical properties of plutonium(IV) and iron(III), siderophore-inspired multidentate hydroxypyridonate ligands have been designed and are unrivaled in terms of actinide-affinity, selectivity, and efficiency. A perspective on the preclinical development of two hydroxypyridonate actinide decorporation agents, 3,4,3-LI(1,2-HOPO) and 5-LIO(Me-3,2-HOPO), is presented. The chemical syntheses of both candidate compounds have been optimized for scale-up. Baseline preparation and analytical methods suitable for manufacturing large amounts have been established. Both ligands show much higher actinide-removal efficacy than the currently approved agent, diethylenetriaminepentaacetic acid (DTPA), with different selectivity for the tested isotopes of plutonium, americium, uranium and neptunium. No toxicity is observed in cells derived from three different human tissue sources treated in vitro up to ligand concentrations of 1 mM, and both ligands were well tolerated in rats when orally administered daily at high doses (>100 micromol kg d) over 28 d under good laboratory practice guidelines. Both compounds are on an accelerated development pathway towards clinical use.

Durbin, Patricia W.; Kullgren, Birgitta; Ebbe, Shirley N.; Xu, Jide; Chang, Polly Y.; Bunin, Deborah I.; Blakely, Eleanor A.; Bjornstad, Kathleen A.; Rosen, Chris J.; Shuh, David K.; Raymond, Kenneth N.

2011-07-13T23:59:59.000Z

333

Toward laser ablation Accelerator Mass Spectrometry of actinides  

Science Conference Proceedings (OSTI)

A project to measure neutron capture cross sections of a number of actinides in a reactor environment by Accelerator Mass Spectrometry (AMS) at the ATLAS facility of Argonne National Laboratory is underway. This project will require the precise and accurate measurement of produced actinide isotopes in many (>30) samples irradiated in the Advanced Test Reactor at Idaho National Laboratory with neutron fluxes having different energy distributions. The AMS technique at ATLAS is based on production of highlycharged positive ions in an electron cyclotron resonance (ECR) ion source followed by acceleration in the ATLAS linac and mass-to-charge (m/q) measurement at the focus of the Fragment Mass Analyzer. Laser ablation was selected as the method of feeding the actinide material into the ion source because we expect it will have higher efficiency and lower chamber contamination than either the oven or sputtering techniques, because of a much narrower angular distribution of emitted material. In addition, a new multi-sample holder/changer to allow quick change between samples and a computer-controlled routine allowing fast tuning of the accelerator for different beams, are being developed. An initial test run studying backgrounds, detector response, and accelerator scaling repeatability was conducted in December 2010. The project design, schedule, and results of the initial test run to study backgrounds are discussed.

R. C. Pardo; F. G. Kondev; S. Kondrashev; C. Nair; T. Palchan; R. Scott; D. Seweryniak; R. Vondrasek; M. Paul; P. Collon; C. Deibel; M. Salvatores; G. Palmiotti; J. Berg; J. Fonnesbeck; G. Imel

2013-01-01T23:59:59.000Z

334

Disposition of actinides released from high-level waste glass  

SciTech Connect

A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90{degrees}C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials.

Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

1994-05-01T23:59:59.000Z

335

Transuranic actinide reactions with simple gas-phase molecules.  

DOE Green Energy (OSTI)

The intent of this research is to conduct an experimental study of f-element chemistry fo r the purpose of identifying reaction trends and mechanisms of the early actinide metals with simple gas phase molecules . Previous research has elucidated some of the fundamenta l chemistry of the 4f elements,1-5 however, more complex chemistry is expected for the 5f serie s due to the inclusion of the 5f electrons in the valence shell . The matrix isolation approach, which is well-suited to the experimental study of transient species, will be used for sample collection, and IR/NIR/VIS spectroscopy will be employed to interrogate deposited matrices . The strength of this method lies in the use of isotopes of reactants, which permits the identification of guest molecules in a noble gas matrix by observation of vibrational frequenc y shifts and patterns upon isotopic substitution . Using this technique at the University of Virginia, the first noble gas-actinide bond has recently been identified, a weak U-Ar bond on the CUO molecule.6 Uranium has similarly been observed to bond to krypton and xenon, whereas thoriu m and the lanthanides have not exhibited this activity . It is expected that plutonium will be even more reactive in this respect . We will extend the body of actinide experimental evidence t o include the transuranic elements neptunium, plutonium, and americium reacted with isotopes o f oxygen, nitrogen, hydrogen, carbon monoxide, and carbon dioxide .

Willson, S. P. (Stephen P.); Veirs, D. K. (Douglas Kirk); Baiardo, J. P. (Joseph P.)

2003-01-01T23:59:59.000Z

336

Analysis of Advanced Actinide-Fueled Energy Systems for Deep Space Propulsion Applications.  

E-Print Network (OSTI)

??The present study is focused on evaluating higher actinides beyond uranium that are capable of supporting power and propulsion requirements in robotic deep space and… (more)

Guy, Troy Lamar

2011-01-01T23:59:59.000Z

337

EA-1404: Actinide Chemistry and Repository Science Laboratory, Carlsbad, New Mexico  

Energy.gov (U.S. Department of Energy (DOE))

This EA evaluates the environmental impacts for the proposal to construct and operate an Actinide Chemistry and Repository Science Laboratory to support chemical research activities related to the...

338

An Experimental Study of Chemical Oxygen Demand Removal from ...  

Science Conference Proceedings (OSTI)

The experimental results showed that the refractory organics in coking wastewater can be effectively removed by this process, and COD removal efficiency was ...

339

Understanding the Chemistry of the Actinides in HL Waste Tank Systems: Actinide Speciation in Oxalic Acid Solutions in the Presence of Significant Quantities of Aluminum, Iron, and Manganese  

SciTech Connect

The overall goal of this research plan is to provide a thermodynamic basis for describing actinide speciation over a range of tank-like conditions, including elevated temperature, elevated OH- concentrations, and the presence of various organic ligands. With support from DOE?s EMSP program, we have made significant progress towards measuring thermodynamic parameters for actinide complexation as a function of temperature. We have used the needs of the ESP modelers to guide our work to date, and we have made important progress defining the effect of temperature for actinide complexation by organic, and for hydrolysis of the hexa- and pentvalent oxidation states.

Clark, Sue

2006-07-30T23:59:59.000Z

340

Fabrication of advanced oxide fuels containing minor actinide for use in fast reactors  

Science Conference Proceedings (OSTI)

R and D of advanced fuel containing minor actinide for use in fast reactors is described related to the composite fuel with MgO matrix. Fabrication tests of MgO composite fuels containing Am were done by a practical process that could be adapted to the presently used commercial manufacturing technology. Am-containing MgO composite fuels having good characteristics, i.e., having no defects, a high density, a homogeneous dispersion of host phase, were obtained. As related technology, burn-up characteristics of a fast reactor core loaded with the present MgO composite fuel were also analyzed, mainly in terms of core criticality. Furthermore, phase relations of MA oxide which was assumed to be contained in MgO matrix fuel were experimentally investigated. (authors)

Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Oarai-machi, Higashi-ibaraki-gun, Ibaraki, 311-1393 (Japan)

2007-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

One of the Largest Pieces of Processing Equipment Removed from Plutonium Finishing Plant- Worker involvement led to safe completion of high-hazard work  

Energy.gov (U.S. Department of Energy (DOE))

RICHLAND, WASH. – U.S. Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CH2M HILL) announced today the successful removal of one of the largest, most complex pieces of equipment from the Plutonium Finishing Plant (PFP) at the Hanford Site in southeast Washington State.

342

Geothermal hydrogen sulfide removal  

DOE Green Energy (OSTI)

UOP Sulfox technology successfully removed 500 ppM hydrogen sulfide from simulated mixed phase geothermal waters. The Sulfox process involves air oxidation of hydrogen sulfide using a fixed catalyst bed. The catalyst activity remained stable throughout the life of the program. The product stream composition was selected by controlling pH; low pH favored elemental sulfur, while high pH favored water soluble sulfate and thiosulfate. Operation with liquid water present assured full catalytic activity. Dissolved salts reduced catalyst activity somewhat. Application of Sulfox technology to geothermal waters resulted in a straightforward process. There were no requirements for auxiliary processes such as a chemical plant. Application of the process to various types of geothermal waters is discussed and plans for a field test pilot plant and a schedule for commercialization are outlined.

Urban, P.

1981-04-01T23:59:59.000Z

343

Contaminant-Organic Complexes: Their Structure and Energetics in Surface Decontamination Processes  

SciTech Connect

The current debate over possible decontamination processes for U.S. Department of Energy (DOE) facilities is centered on disparate decontamination problems, but the key contaminants (uranium [U], plutonium [Pu], and neptunium [Np]) are universally important. There is no single decontamination technique or agent for all metal surfaces and contaminants with which DOE is faced. However, more innovative agents used alone or in conjunction with traditional processes can increase the potential to reclaim for future use some of these valuable resources or, at the least, decontaminate the metal surfaces to allow disposal as nonradioactive, nonhazardous material. This debate underscores several important issues: (1) regardless of the decontamination scenario, metal (Fe, U, Pu, Np) oxide film removal from the surface is central to decontamination; and (2) simultaneous oxide dissolution and sequestration of actinide contaminants against re-adsorption to a clean metal surface will influence the efficacy of a process or agent and its cost.

Ainsworth, Calvin C.; Hay, Benjamin P.; Traina, Samuel J.; Myneni, Satish C. B.

2002-06-01T23:59:59.000Z

344

LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium  

Science Conference Proceedings (OSTI)

Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and interactions occurring within the plasma, such as collisional energy transfer, that might be a factor in the reduction in neptunium emission lines. Neptunium has to be analyzed alone using LIBS to further understand the dynamics that may be occurring in the plasma of the mixed actinide fuel pellet sample. The LIBS data suggests that the emission spectrum for the mixed actinide fuel pellet is not simply the sum of the emission spectra of the pure samples but is dependent on the species present in the plasma and the interactions and reactions that occur within the plasma. Finally, many of the neptunium lines are in the near infrared region which is drastically reduced in intensity by the current optical setup and possibly the sensitivity of the emission detector in the spectral region. Once the optics are replaced and the optical collection system is modified and optimized, the probability of observing emission lines for neptunium might be increased significantly. The mixed actinide fuel pellet was analyzed under the experimental conditions listed in Table 1. The LIBS spectra of the fuel pellet are shown in Figures 1-49. The spectra are labeled with the observed wavelength and atomic species (both neutral (I) and ionic (II)). Table 2 is a complete list of the observed and literature based emission wavelengths. The literature wavelengths have references including NIST Atomic Spectra Database (NIST), B.A. Palmer et al. 'An Atlas of Uranium Emission Intensities in a Hollow Cathode Discharge' taken at the Kitt Peak National Observatory (KPNO), R.L. Kurucz 1995 Atomic Line Data from the Smithsonian Astrophysical Observatory (SAO), J. Blaise et al. 'The Atomic Spectrum of Plutonium' from Argonne National Laboratory (BFG), and M. Fred and F.S. Tomkins, 'Preliminary Term Analysis of Am I and Am II Spectra' (FT). The dash (-) shown under Ionic State indicates that the ionic state of the transition was not available. In the spectra, the dash (-) is replaced with a question mark (?). Peaks that are not assigned are most likely real features and not noise but cannot be confidently assi

Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

2012-06-18T23:59:59.000Z

345

Final Report on Actinide Glass Scintillators for Fast Neutron Detection  

Science Conference Proceedings (OSTI)

This is the final report of an experimental investigation of actinide glass scintillators for fast-neutron detection. It covers work performed during FY2012. This supplements a previous report, PNNL-20854 “Initial Characterization of Thorium-loaded Glasses for Fast Neutron Detection” (October 2011). The work in FY2012 was done with funding remaining from FY2011. As noted in PNNL-20854, the glasses tested prior to July 2011 were erroneously identified as scintillators. The decision was then made to start from “scratch” with a literature survey and some test melts with a non-radioactive glass composition that could later be fabricated with select actinides, most likely thorium. The normal stand-in for thorium in radioactive waste glasses is cerium in the same oxidation state. Since cerium in the 3+ state is used as the light emitter in many scintillating glasses, the next most common substitute was used: hafnium. Three hafnium glasses were melted. Two melts were colored amber and a third was clear. It barely scintillated when exposed to alpha particles. The uses and applications for a scintillating fast neutron detector are important enough that the search for such a material should not be totally abandoned. This current effort focused on actinides that have very high neutron capture energy releases but low neutron capture cross sections. This results in very long counting times and poor signal to noise when working with sealed sources. These materials are best for high flux applications and access to neutron generators or reactors would enable better test scenarios. The total energy of the neutron capture reaction is not the only factor to focus on in isotope selection. Many neutron capture reactions result in energetic gamma rays that require large volumes or high densities to detect. If the scintillator is to separate neutrons from gamma rays, the capture reactions should produce heavy particles and few gamma rays. This would improve the detection of a signal for fast neutron capture.

Bliss, Mary; Stave, Jean A.

2012-10-01T23:59:59.000Z

346

Plutonium and minor actinides utilization in Thorium molten salt reactor  

Science Conference Proceedings (OSTI)

FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/{sup 233}U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu and MA composition in the fuel of 5.96% or more.

Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

347

Removal of radioactive materials and heavy metals from water using magnetic resin  

DOE Patents (OSTI)

Magnetic polymer resins capable of efficient removal of actinides and heavy metals from contaminated water are disclosed together with methods for making, using, and regenerating them. The resins comprise polyamine-epichlorohydrin resin beads with ferrites attached to the surfaces of the beads. Markedly improved water decontamination is demonstrated using these magnetic polymer resins of the invention in the presence of a magnetic field, as compared with water decontamination methods employing ordinary ion exchange resins or ferrites taken separately.

Kochen, Robert L. (Boulder, CO); Navratil, James D. (Simi Valley, CA)

1997-01-21T23:59:59.000Z

348

Microbial Transformation of TRU and Mixed Waste: Actinide Speciation and Waste Volume  

Science Conference Proceedings (OSTI)

In order to understand the susceptibility of transuranic and mixed waste to microbial degradation (as well as any mechanism which depends upon either complexation and/or redox of metal ions), it is essential to understand the association of metal ions with organic ligands present in mixed wastes. These ligands have been found in our previous EMSP study to limit electron transfer reactions and strongly affect transport and the eventual fate of radionuclides in the environment. As transuranic waste (and especially mixed waste) will be retained in burial sites and in legacy containment for (potentially) many years while awaiting treatment and removal (or remaining in place under stewardship agreements at government subsurface waste sites), it is also essential to understand the aging of mixed wastes and its implications for remediation and fate of radionuclides. Mixed waste containing actinides and organic materials are especially complex and require extensive study. The EMSP program described in this report is part of a joint program with the Environmental Sciences Department at Brookhaven National Laboratory. The Stony Brook University portion of this award has focused on the association of uranium (U(VI)) and transuranic analogs (Ce(III) and Eu(III)) with cellulosic materials and related compounds, with development of implications for microbial transformation of mixed wastes. The elucidation of the chemical nature of mixed waste is essential for the formulation of remediation and encapsulation technologies, for understanding the fate of contaminant exposed to the environment, and for development of meaningful models for contaminant storage and recovery.

Halada, Gary P

2008-04-10T23:59:59.000Z

349

Process analyses and monitoring system in labview data logging and supervisory control module of the cryogenic pilot plant for tritium removal  

Science Conference Proceedings (OSTI)

The system responds to the monitoring requirements of the technological processes specific to the nuclear installation that processes radioactive substances, with severe consequences in case of technological failure, as is the case with a tritium processing ... Keywords: monitoring system, process analyses

Carmen Maria Moraru; Iuliana Stefan; Ovidiu Balteanu; Ciprian Bucur; Liviu Stefan; Anisia Bornea; Ioan Stefanescu

2008-02-01T23:59:59.000Z

350

Effects of actinide burning on waste disposal at Yucca Mountain  

SciTech Connect

Release rates of 15 radionuclides from waste packages expected to result from partitioning and transmutation of Light-Water Reactor (LWR) and Actinide-Burning Liquid-Metal Reactor (ALMR) spent fuel are calculated and compared to release rates from standard LWR spent fuel packages. The release rates are input to a model for radionuclide transport from the proposed geologic repository at Yucca Mountain to the water table. Discharge rates at the water table are calculated and used in a model for transport to the accessible environment, defined to be five kilometers from the repository edge. Concentrations and dose rates at the accessible environment from spent fuel and wastes from reprocessing, with partitioning and transmutation, are calculated. Partitioning and transmutation of LWR and ALMR spent fuel reduces the inventories of uranium, neptunium, plutonium, americium and curium in the high-level waste by factors of 40 to 500. However, because release rates of all of the actinides except curium are limited by solubility and are independent of package inventory, they are not reduced correspondingly. Only for curium is the repository release rate much lower for reprocessing wastes.

Hirschfelder, J. [California Univ., Berkeley, CA (United States)

1992-07-01T23:59:59.000Z

351

Conservative axial burnup distributions for actinide-only burnup credit  

SciTech Connect

Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit.

Kang, C.; Lancaster, D.

1997-11-01T23:59:59.000Z

352

Salt Processing at the Savannah River Site: Results of Technology Down-Selection and Research and Development to Support New Salt Waste Processing Facility  

Science Conference Proceedings (OSTI)

The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste (HLW) program is responsible for storage, treatment, and immobilization of HLW for disposal. The Salt Processing Project (SPP) is the salt waste (water-soluble) treatment portion of this effort. The overall SPP encompasses the selection, design, construction, and operation of technologies to prepare the salt-waste feed material for immobilization at the site's Saltstone Production Facility (SPF) and vitrification facility (Defense Waste Processing Facility [DWPF]). Major constituents that must be removed from the salt waste and sent as feed to DWPF include cesium (Cs), strontium (Sr), and actinides. In April 2000, the DOE Deputy Secretary for Project Completion (EM-40) established the SRS Salt Processing Project Technical Working Group (TWG) to manage technology development of treatment alternatives for SRS high-level salt wastes. The separation alternatives investigated included three candidate Cs-removal processes selected, as well as actinide and Sr removal that are also required as a part of each process. The candidate Cs-removal processes are: crystalline Silicotitanate Non-Elutable Ion Exchange (CST); caustic Side Solvent Extraction (CSSX); and small Tank Tetraphenylborate Precipitation (STTP). The Tanks Focus Area was asked to assist DOE by managing the SPP research and development (R&D), revising roadmaps, and developing down-selection criteria. The down-selection decision process focused its analysis on three levels: (a) identification of goals that the selected technology should achieve, (b) selection criteria that are a measure of performance of the goal, and (c) criteria scoring and weighting for each technology alternative. After identifying the goals and criteria, the TWG analyzed R&D results and engineering data and scored the technology alternatives versus the criteria. Based their analysis and scoring, the TWG recommended CSSX as the preferred alternative. This recommendation was formalized in July 2001 when DOE published the Savannah River Site Salt Processing Alternatives Final Supplemental Environmental Impact Statement (SEIS) and was finalized in the DOE Record of Decision issued in October 2001.

Lang, K.; Gerdes, K.; Picha, K.; Spader, W.; McCullough, J.; Reynolds, J.; Morin, J. P.; Harmon, H. D.

2002-02-26T23:59:59.000Z

353

Literature review of United States utilities computer codes for calculating actinide isotope content in irradiated fuel  

SciTech Connect

This paper reviews the accuracy and precision of methods used by United States electric utilities to determine the actinide isotopic and element content of irradiated fuel. After an extensive literature search, three key code suites were selected for review. Two suites of computer codes, CASMO and ARMP, are used for reactor physics calculations; the ORIGEN code is used for spent fuel calculations. They are also the most widely used codes in the nuclear industry throughout the world. Although none of these codes calculate actinide isotopics as their primary variables intended for safeguards applications, accurate calculation of actinide isotopic content is necessary to fulfill their function.

Horak, W.C.; Lu, Ming-Shih

1991-12-01T23:59:59.000Z

354

FY2011 Annual Report for the Actinide Isomer Detection Project  

Science Conference Proceedings (OSTI)

This project seeks to identify a new signature for actinide element detection in active interrogation. This technique works by exciting and identifying long-lived nuclear excited states (isomers) in the actinide isotopes and/or primary fission products. Observation of isomers in the fission products will provide a signature for fissile material. For the actinide isomers, the decay time and energy of the isomeric state is unique to a particular isotope, providing an unambiguous signature for SNM. This project entails isomer identification and characterization and neutron population studies. This document summarizes activities from its third year - completion of the isomer identification characterization experiments and initialization of the neutron population experiments. The population and decay of the isomeric state in 235U remain elusive, although a number of candidate gamma rays have been identified. In the course of the experiments, a number of fission fragment isomers were populated and measured [Ressler 2010]. The decays from these isomers may also provide a suitable signature for the presence of fissile material. Several measurements were conducted throughout this project. This report focuses on the results of an experiment conducted collaboratively by PNNL, LLNL and LBNL in December 2010 at LBNL. The measurement involved measuring the gamma-rays emitted from an HEU target when bombarded with 11 MeV neutrons. This report discussed the analysis and resulting conclusions from those measurements. There was one strong candidate, at 1204 keV, of an isomeric signature of 235U. The half-life of the state is estimated to be 9.3 {mu}s. The measured time dependence fits the decay time structure very well. Other possible explanations for the 1204-keV state were investigated, but they could not explain the gamma ray. Unfortunately, the relatively limited statistics of the measurement limit, and the lack of understanding of some of the systematic of the experiment, limit the authors to labeling the 1204-keV gamma ray as a very strong candidate for isomeric transition in 235U. Regardless of the physics origins, the time structure of the 1204-keV gamma ray can be used as at a minimum as an indication of fissile material, if the 1204-keV gamma ray is attributed to a fission product, or it may be a unique signature for 235U, if it is a signature of an isomeric state in 235U.

Warren, Glen A.; Francy, Christopher J.; Ressler, Jennifer J.; Erikson, Luke E.; Tatishvili, Gocha; Hatarik, R.

2011-10-01T23:59:59.000Z

355

Regenerative process for removal of mercury and other heavy metals from gases containing H.sub.2 and/or CO  

DOE Patents (OSTI)

A method for removal of mercury from a gaseous stream containing the mercury, hydrogen and/or CO, and hydrogen sulfide and/or carbonyl sulfide in which a dispersed Cu-containing sorbent is contacted with the gaseous stream at a temperature in the range of about 25.degree. C. to about 300.degree. C. until the sorbent is spent. The spent sorbent is contacted with a desorbing gaseous stream at a temperature equal to or higher than the temperature at which the mercury adsorption is carried out, producing a regenerated sorbent and an exhaust gas comprising released mercury. The released mercury in the exhaust gas is captured using a high-capacity sorbent, such as sulfur-impregnated activated carbon, at a temperature less than about 100.degree. C. The regenerated sorbent may then be used to capture additional mercury from the mercury-containing gaseous stream.

Jadhav, Raja A. (Naperville, IL)

2009-07-07T23:59:59.000Z

356

Fission Cross Section Measurements of Actinides at LANSCE  

Science Conference Proceedings (OSTI)

Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications. By combining measurement at two LANSCE facilities, Lujan Center and the Weapons Neutron Research center (WNR), differential cross sections can be measured from sub-thermal energies up to 200 MeV. Incident neutron energies are determined using the time-of-flight method, and parallel-plate ionization chambers are used to measure fission cross sections relative to the 235U standard. Recent measurements include the 233, 238U, 239-242Pu, and 243Am neutron-induced fission cross sections. In this paper preliminary results for fission cross sections of 243Am and 233U will be presented.

F. Tovesson; A. B. Laptev; T. S. Hill

2011-08-01T23:59:59.000Z

357

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers  

SciTech Connect

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

2013-03-21T23:59:59.000Z

358

Delayed neutron measurements from fast fission of actinide waste isotopes  

E-Print Network (OSTI)

A study was performed to determine the delayed neutron emission properties from fast fission of several actinide waste isotopes. The specific isotopes evaluated were U-235, Np-237, and Am-243. A calculational technique based on the microscopic method was used to predict initial guesses for the delayed neutron parameters (group decay constants and yields). Based on these calculations, an alternate "seven-group" structure, in contrast to the traditional "six-group" structure used previously, was suggested which would yield a superior fit to the measured data. A series of measurements were performed to test the hypothesis suggested by this alternate group structure. Using a set of highly purified actinide samples (provided by Oak Ridge National Laboratory), the delayed neutron emission decay constants and yields for six groups of the "seven-group" structure were measured for U-235, Np-237, and Am-243. These experiments were performed using the Texas A&M University Nuclear Science Center Reactor, a quick pneumatic transfer system, an integrated computer control and counting system, and a specially designed in-core irradiation device. The values for the total delayed neutron yield (per 100 fissions) from fast-neutron induced fission of U-235, Np237, and Am-243 were determined to be 1.67 ?0.08, 1.14 ?0.07, 0.86 ?0.05, respectively. The newly measured values were compared with other values recommended by Keepin et al., Waldo et al., Saleh et al., and Brady and England. Good agreement was found in all cases. The "seven-group" structure was shown to yield a superior fit to the measured data, as well as, provide a more direct correlation between delayed neutron groups and their associated delayed neutron precursors.

Charlton, William S.

1997-01-01T23:59:59.000Z

359

Fission of actinides and superheavy nuclei: covariant density functional theory perspective  

E-Print Network (OSTI)

The current status of the application of covariant density functional theory to the description of fission barriers in actinides and superheavy nuclei is reviewed. The achievements and open problems are discussed.

A. V. Afanasjev

2013-03-05T23:59:59.000Z

360

Analysis of the MIT research reactor fission product and actinide radioactivity inventories  

E-Print Network (OSTI)

The current analysis of the MITR core radioactivity inventory eliminates unnecessary assumptions made in previous estimates of the inventory, and revises the list of contributory isotopes to include all actinide and fission ...

Kennedy, William B. (William Blake), 1979-

2004-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

In situ removal of contamination from soil  

DOE Patents (OSTI)

A process of remediation of cationic heavy metal contamination from soil utilizes gas phase manipulation to inhibit biodegradation of a chelating agent that is used in an electrokinesis process to remove the contamination, and further gas phase manipulation to stimulate biodegradation of the chelating agent after the contamination has been removed. The process ensures that the chelating agent is not attacked by bioorganisms in the soil prior to removal of the contamination, and that the chelating agent does not remain as a new contaminant after the process is completed.

Lindgren, Eric R. (Albuquerque, NM); Brady, Patrick V. (Albuquerque, NM)

1997-01-01T23:59:59.000Z

362

Large Component Removal/Disposal  

Science Conference Proceedings (OSTI)

This paper describes the removal and disposal of the large components from Maine Yankee Atomic Power Plant. The large components discussed include the three steam generators, pressurizer, and reactor pressure vessel. Two separate Exemption Requests, which included radiological characterizations, shielding evaluations, structural evaluations and transportation plans, were prepared and issued to the DOT for approval to ship these components; the first was for the three steam generators and one pressurizer, the second was for the reactor pressure vessel. Both Exemption Requests were submitted to the DOT in November 1999. The DOT approved the Exemption Requests in May and July of 2000, respectively. The steam generators and pressurizer have been removed from Maine Yankee and shipped to the processing facility. They were removed from Maine Yankee's Containment Building, loaded onto specially designed skid assemblies, transported onto two separate barges, tied down to the barges, th en shipped 2750 miles to Memphis, Tennessee for processing. The Reactor Pressure Vessel Removal Project is currently under way and scheduled to be completed by Fall of 2002. The planning, preparation and removal of these large components has required extensive efforts in planning and implementation on the part of all parties involved.

Wheeler, D. M.

2002-02-27T23:59:59.000Z

363

High removal rate laser-based coating removal system  

DOE Patents (OSTI)

A compact laser system that removes surface coatings (such as paint, dirt, etc.) at a removal rate as high as 1000 ft.sup.2 /hr or more without damaging the surface. A high repetition rate laser with multiple amplification passes propagating through at least one optical amplifier is used, along with a delivery system consisting of a telescoping and articulating tube which also contains an evacuation system for simultaneously sweeping up the debris produced in the process. The amplified beam can be converted to an output beam by passively switching the polarization of at least one amplified beam. The system also has a personal safety system which protects against accidental exposures.

Matthews, Dennis L. (Moss Beach, CA); Celliers, Peter M. (Berkeley, CA); Hackel, Lloyd (Livermore, CA); Da Silva, Luiz B. (Danville, CA); Dane, C. Brent (Livermore, CA); Mrowka, Stanley (Richmond, CA)

1999-11-16T23:59:59.000Z

364

Turbomachinery debris remover  

DOE Patents (OSTI)

An apparatus for removing debris from a turbomachine. The apparatus includes housing and remotely operable viewing and grappling mechanisms for the purpose of locating and removing debris lodged between adjacent blades in a turbomachine.

Krawiec, Donald F. (Pittsburgh, PA); Kraf, Robert J. (North Huntingdon, PA); Houser, Robert J. (Monroeville, PA)

1988-01-01T23:59:59.000Z

365

The Joint Actinide Shock Physics Experimental Research Facility at the Nevada National Security Site, OAS-L-12-05  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Joint Actinide Shock Physics Joint Actinide Shock Physics Experimental Research Facility at the Nevada National Security Site OAS-L-12-05 April 2012 Department of Energy Washington, DC 20585 April 23, 2012 MEMORANDUM FOR THE MANAGER, NEVADA SITE OFFICE FROM: David Sedillo, Director Western Audits Division Office of Inspector General SUBJECT: INFORMATION: Audit Report on "The Joint Actinide Shock Physics Experimental Research Facility at the Nevada National Security Site" BACKGROUND The Department of Energy, National Nuclear Security Administration's, Joint Actinide Shock Physics Experimental Research (JASPER) facility plays an integral role in the certification of the Nation's nuclear weapons stockpile by providing a method to generate and measure data

366

Enhancing BWR Proliferation Resistance Fuel with Minor Actinides  

SciTech Connect

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

Gray S. Chang

2009-03-01T23:59:59.000Z

367

Proliferation Resistance Evaluation of ACR-1000 Fuel with Minor Actinides  

Science Conference Proceedings (OSTI)

The Global Nuclear Energy Partnership (GNEP) program is to significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance

Gray S. Chang

2008-09-01T23:59:59.000Z

368

Enhancing VVER Annular Proliferation Resistance Fuel with Minor Actinides  

SciTech Connect

Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. The merits of nuclear energy are the high-density energy, and low environmental impacts i.e. almost zero greenhouse gas emission. Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current LWR as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce the spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope 238Pu /Pu ratio. For future advanced nuclear systems, the minor actinides are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. In this paper, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. We concluded that the concept of MARA, involves the use of transuranic nuclides (237Np and/or 241Am), can not only drastically increase the 238Pu/Pu ratio for proliferation resistance, but also can serve as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

G. S. Chang

2007-06-01T23:59:59.000Z

369

Graphitic packing removal tool  

DOE Patents (OSTI)

Graphitic packing removal tools are described for removal of the seal rings in one piece from valves and pumps. The packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

Meyers, K.E.; Kolsun, G.J.

1996-12-31T23:59:59.000Z

370

Actinide production in /sup 136/Xe bombardments of /sup 249/Cf  

Science Conference Proceedings (OSTI)

The production cross sections for the actinide products from /sup 136/Xe bombardments of /sup 249/Cf at energies 1.02, 1.09, and 1.16 times the Coulomb barrier were determined. Fractions of the individual actinide elements were chemically separated from recoil catcher foils. The production cross sections of the actinide products were determined by measuring the radiations emitted from the nuclides within the chemical fractions. The chemical separation techniques used in this work are described in detail, and a description of the data analysis procedure is included. The actinide production cross section distributions from these /sup 136/Xe + /sup 249/Cf bombardments are compared with the production cross section distributions from other heavy ion bombardments of actinide targets, with emphasis on the comparison with the /sup 136/Xe + /sup 248/Cm reaction. A technique for modeling the final actinide cross section distributions has been developed and is presented. In this model, the initial (before deexcitation) cross section distribution with respect to the separation energy of a dinuclear complex and with respect to the Z of the target-like fragment is given by an empirical procedure. It is then assumed that the N/Z equilibration in the dinuclear complex occurs by the transfer of neutrons between the two participants in the dinuclear complex. The neutrons and the excitation energy are statistically distributed between the two fragments using a simple Fermi gas level density formalism. The resulting target-like fragment initial cross section distribution with respect to Z, N, and excitation energy is then allowed to deexcite by emission of neutrons in competition with fission. The result is a final cross section distribution with respect to Z and N for the actinide products. 68 refs., 33 figs., 6 tabs.

Gregorich, K.E.

1985-08-01T23:59:59.000Z

371

Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials  

SciTech Connect

Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO{sub 2} and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for {sup 237}Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO{sub 2}, and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity.

Becker, Udo [Univ. of Michigan (United States)

2013-04-16T23:59:59.000Z

372

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

2013-10-01T23:59:59.000Z

373

Actinide recycle potential in the IFR (Integral Fast Reactor)  

SciTech Connect

Rising concern about the greenhouse effect reinforces the need to reexamine the question of a next-generation reactor concept that can contribute significantly toward substitution for fossil-based energy generation. Even with only the nuclear capacity on-line today, world-wide reasonably assured uranium resources would last for only about 50 years. If nuclear is to make a significant contribution, breeding is a fundamental requirement. Uranium resources can then be extended by two orders of magnitude, making nuclear essentially a renewable energy source. The key technical elements of the IFR concept are metallic fuel and fuel cycle technology based on pyroprocessing. Pyroprocessing is radically different from the conventional PUREX reprocessing developed for the LWR oxide fuel. Chemical feasibility of pyroprocessing has been demonstrated. The next major step in the IFR development program will be the full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. IFR fuel cycle closure based on pyroprocessing can also have a dramatic impact on the waste management options, and in particular on the actinide recycling. 6 figs.

Chang, Y.I.

1989-01-01T23:59:59.000Z

374

Partition of actinides and fission products between metal and molten salt phases: Theory, measurement, and application to IFR pyroprocess development  

Science Conference Proceedings (OSTI)

The chemical basis of Integral Fast Reactor fuel reprocessing (pyroprocessing) is partition of fuel, cladding, and fission product elements between molten LiCl-KCl and either a solid metal phase or a liquid cadmium phase. The partition reactions are described herein, and the thermodynamic basis for predicting distributions of actinides and fission products in the pyroprocess is discussed. The critical role of metal-phase activity coefficients, especially those of rare earth and the transuranic elements, is described. Measured separation factors, which are analogous to equilibrium constants but which involve concentrations rather than activities, are presented. The uses of thermodynamic calculations in process development are described, as are computer codes developed for calculating material flows and phase compositions in pyroprocessing.

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

1993-10-01T23:59:59.000Z

375

Procession  

E-Print Network (OSTI)

UEE 2008 Ziermann, Martin 2004 Macht und Architektur: ZweiP ROCESSION Martin Stadler EDITORS W ILLEKE W ENDRICHFull Citation: Stadler, Martin, 2008, Procession. In Jacco

Stadler, Martin

2008-01-01T23:59:59.000Z

376

Processing  

Science Conference Proceedings (OSTI)

...are processed to complex final shapes by investment casting. Iron-nickel-base superalloys are not customarily investment cast. Investment casting permits intricate internal cooling

377

Fluorination process using catalysts  

DOE Patents (OSTI)

A process is given for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/, AgF/sub 2/ and NiF/sub 2/, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/ and AgF/sub 2/, whereby the fluorination is significantly enhanced.

Hochel, R.C.; Saturday, K.A.

1983-08-25T23:59:59.000Z

378

Fluorination process using catalyst  

DOE Patents (OSTI)

A process for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3, AgF.sub.2 and NiF.sub.2, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3 and AgF.sub.2, whereby the fluorination is significantly enhanced.

Hochel, Robert C. (Aiken, SC); Saturday, Kathy A. (Aiken, SC)

1985-01-01T23:59:59.000Z

379

Aqueous recovery of plutonium from pyrochemical processing residues  

Science Conference Proceedings (OSTI)

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

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

1984-01-01T23:59:59.000Z

380

Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor  

SciTech Connect

This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "actinide removal process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Technical requirements for the actinide source-term waste test program  

SciTech Connect

This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency`s 40 CFR Part 191B.

Phillips, M.L.F.; Molecke, M.A.

1993-10-01T23:59:59.000Z

382

A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal  

SciTech Connect

The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12 x 10-6 g?m-2 and 2.7 x 10-3 g?m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.

Donald, Ian W.; Metcalfe, Brian; Fong, Shirley K.; Gerrard, Lee A.; Strachan, Denis M.; Scheele, Randall D.

2007-03-31T23:59:59.000Z

383

Enhancing BWR Proliferation Resistance Fuel with Minor Actinides  

Science Conference Proceedings (OSTI)

Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. We concluded that the concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy rennaissance.

Gray S. Chang

2008-07-01T23:59:59.000Z

384

Final Project Report INERT-MATRIX FUEL: ACTINIDE "BURNING" AND DIRECT DISPOSAL  

Office of Scientific and Technical Information (OSTI)

Project Report Project Report INERT-MATRIX FUEL: ACTINIDE "BURNING" AND DIRECT DISPOSAL Nuclear Engineering Education Research Program (grant # DE-FG07-99ID13767) Rodney C. Ewing (co-PI) Lumin Wang (co-PI) October 30,2002 For the Period of 07/01/1999 to 06/30/2002 Department of Nuclear Engineering and Radiological Sciences University of Michigan Ann Arbor, MI 48109 1 1. Background Excess actinides result from the dismantlement of nuclear weapons (239Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241Am, Cm and 237Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burn- up of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-

385

Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR  

Science Conference Proceedings (OSTI)

One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

G. S. Chang; Hongbin Zhang

2009-09-01T23:59:59.000Z

386

Device for removing blackheads  

DOE Patents (OSTI)

A device for removing blackheads from pores in the skin having a elongated handle with a spoon shaped portion mounted on one end thereof, the spoon having multiple small holes piercing therethrough. Also covered is method for using the device to remove blackheads.

Berkovich, Tamara (116 N. Wetherly Dr., Suite 115, Los Angeles, CA)

1995-03-07T23:59:59.000Z

387

Some activities in the United States concerning the physics aspects of actinide waste recycling  

SciTech Connect

Reactor types being considered in the United States for the purpose of actinide waste recycling are discussed briefly. The reactor types include thermal reactors operating on the 3.3 percent $sup 235$U--$sup 238$U and the $sup 233$U--$sup 232$Th fuel cycles, liquid metal fast breeder reactors, reactors fueled entirely by actinide wastes, gaseous fuel reactors, and fusion reactors. Cross section measurements in progress or planned toward providing basic data for testing the recycle concept are also discussed. (auth)

Raman, S.

1975-01-01T23:59:59.000Z

388

Economic feasibility of biochemical processes for the upgrading of crudes and the removal of sulfur, nitrogen, and trace metals from crude oil -- Benchmark cost establishment of biochemical processes on the basis of conventional downstream technologies. Final report FY95  

Science Conference Proceedings (OSTI)

During the past several years, a considerable amount of work has been carried out showing that microbially enhanced oil recovery (MEOR) is promising and the resulting biotechnology may be deliverable. At Brookhaven National Laboratory (BNL), systematic studies have been conducted which dealt with the effects of thermophilic and thermoadapted bacteria on the chemical and physical properties of selected types of crude oils at elevated temperatures and pressures. Current studies indicate that during the biotreatment several chemical and physical properties of crude oils are affected. The oils are (1) emulsified; (2) acidified; (3) there is a qualitative and quantitative change in light and heavy fractions of the crudes; (4) there are chemical changes in fractions containing sulfur compounds; (5) there is an apparent reduction in the concentration of trace metals; and (6) the qualitative and quantitative changes appear to be microbial species dependent; and (7) there is a distinction between biodegraded and biotreated oils. The downstream biotechnological crude oil processing research performed thus far is of laboratory scale and has focused on demonstrating the technical feasibility of downstream processing with different types of biocatalysts under a variety of processing conditions. Quantitative economic analysis is the topic of the present project which investigates the economic feasibility of the various biochemical downstream processes which hold promise in upgrading of heavy crudes, such as those found in California, e.g., Monterey-type, Midway Sunset, Honda crudes, and others.

Premuzic, E.T.

1996-08-01T23:59:59.000Z

389

Microsoft Word - ARP MCU milestone.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

waste in its Interim Salt Disposition Process (ISDP) at the Savannah River Site (SRS). Essential components of the ISDP is the Actinide Removal Process (ARP) and Modular Caustic...

390

Synthesis and development of processes for the recovery of sulfur from acid gases. Part 1, Development of a high-temperature process for removal of H{sub 2}S from coal gas using limestone -- thermodynamic and kinetic considerations; Part 2, Development of a zero-emissions process for recovery of sulfur from acid gas streams  

SciTech Connect

Limestone can be used more effectively as a sorbent for H{sub 2}S in high-temperature gas-cleaning applications if it is prevented from undergoing calcination. Sorption of H{sub 2}S by limestone is impeded by sintering of the product CaS layer. Sintering of CaS is catalyzed by CO{sub 2}, but is not affected by N{sub 2} or H{sub 2}. The kinetics of CaS sintering was determined for the temperature range 750--900{degrees}C. When hydrogen sulfide is heated above 600{degrees}C in the presence of carbon dioxide elemental sulfur is formed. The rate-limiting step of elemental sulfur formation is thermal decomposition of H{sub 2}S. Part of the hydrogen thereby produced reacts with CO{sub 2}, forming CO via the water-gas-shift reaction. The equilibrium of H{sub 2}S decomposition is therefore shifted to favor the formation of elemental sulfur. The main byproduct is COS, formed by a reaction between CO{sub 2} and H{sub 2}S that is analogous to the water-gas-shift reaction. Smaller amounts of SO{sub 2} and CS{sub 2} also form. Molybdenum disulfide is a strong catalyst for H{sub 2}S decomposition in the presence of CO{sub 2}. A process for recovery of sulfur from H{sub 2}S using this chemistry is as follows: Hydrogen sulfide is heated in a high-temperature reactor in the presence of CO{sub 2} and a suitable catalyst. The primary products of the overall reaction are S{sub 2}, CO, H{sub 2} and H{sub 2}O. Rapid quenching of the reaction mixture to roughly 600{degrees}C prevents loss Of S{sub 2} during cooling. Carbonyl sulfide is removed from the product gas by hydrolysis back to CO{sub 2} and H{sub 2}S. Unreacted CO{sub 2} and H{sub 2}S are removed from the product gas and recycled to the reactor, leaving a gas consisting chiefly of H{sub 2} and CO, which recovers the hydrogen value from the H{sub 2}S. This process is economically favorable compared to the existing sulfur-recovery technology and allows emissions of sulfur-containing gases to be controlled to very low levels.

Towler, G.P.; Lynn, S.

1993-05-01T23:59:59.000Z

391

PROCESS FOR REMOVING NOBLE METALS FROM URANIUM  

DOE Patents (OSTI)

A pyrometallurgical method is given for purifying uranium containing ruthenium and palladium. The uranium is disintegrated and oxidized by exposure to air and then the ruthenium and palladium are extracted from the uranium with molten zinc.

Knighton, J.B.

1961-01-31T23:59:59.000Z

392

Biotechnology Based Processes for Arsenic Removal  

Science Conference Proceedings (OSTI)

Selective Recovery of Gold from E-wastes by Using Cellulosic Wastes · Stabilization of Chromium-Based Slags with FeS2 and FeSO4 · Sulphide Precipitation ...

393

High removal rate laser-based coating removal system  

Science Conference Proceedings (OSTI)

A compact laser system is disclosed that removes surface coatings (such as paint, dirt, etc.) at a removal rate as high as 1,000 ft{sup 2}/hr or more without damaging the surface. A high repetition rate laser with multiple amplification passes propagating through at least one optical amplifier is used, along with a delivery system consisting of a telescoping and articulating tube which also contains an evacuation system for simultaneously sweeping up the debris produced in the process. The amplified beam can be converted to an output beam by passively switching the polarization of at least one amplified beam. The system also has a personal safety system which protects against accidental exposures.

Matthews, D.L.; Celliers, P.M.; Hackel, L.; Da Silva, L.B.; Dane, C.B.; Mrowka, S.

1999-11-16T23:59:59.000Z

394

Addressing mixed waste in plutonium processing  

SciTech Connect

The overall goal is the minimization of all waste generated in actinide processing facilities. Current emphasis is directed toward reducing and managing mixed waste in plutonium processing facilities. More specifically, the focus is on prioritizing plutonium processing technologies for development that will address major problems in mixed waste management. A five step methodological approach to identify, analyze, solve, and initiate corrective action for mixed waste problems in plutonium processing facilities has been developed.

Christensen, D.C.; Sohn, C.L. (Los Alamos National Lab., NM (United States)); Reid, R.A. (New Mexico Univ., Albuquerque, NM (United States). Anderson Schools of Management)

1991-01-01T23:59:59.000Z

395

Virtual Design and Modeling of Various Manufacturing Processes for Remote . . .  

E-Print Network (OSTI)

As currently envisioned, over 70,000 tons of high-level nuclear waste would be stored inside the planned Yucca Mountain repository. After emplacement, the site must be maintained and guarded for over 10,000 years. The reprocessing of spent nuclear fuel is a possible alternative to geological storage. Here, depleted Uranium would be separated from Plutonium and Minor Actinides. While Plutonium can be ‘burned ’ in commercial nuclear reactors, the minor actinides would be transmuted into other elements. The large-scale deployment of remote fabrication and refabrication processes (approx. 100 tons of Minor Actinides (MA) annually) will be required. Process automation has the potential to decrease the cost of remote fuel fabrication and to make transmutation a more economically viable process. Reprocessing and transmutation would reduce the high-level waste volume by over 99%, and reduce the lifetime of the repository to approximately 300 years. The objective of this thesis is the virtual design and simulation of manufacturing

Jamil Mohamad Renno

2005-01-01T23:59:59.000Z

396

Chloride removal from plutonium alloy  

Science Conference Proceedings (OSTI)

SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP.

Holcomb, H.P.

1983-01-01T23:59:59.000Z

397

Countercurrent flowsheet testing of the TRUEX process with ICPP calcine  

SciTech Connect

Calcine was generated at the Idaho Chemical Processing Plant over several decades as a method of solidifying numerous raffinates and wastes from spent nuclear fuel reprocessing for convenient interim storage. Unfortunately, the bulk of the calcine is inert, with radionuclides comprising less than 1 weight percent of the total calcine mass. The bulk of the calcine currently stored at the ICPP was produced from wastes generated during reprocessing of zirconium clad fuels. Consequently, this material contains varying, but large quantities of zirconium oxide. Currently, separations options are being considered for acidic solutions of dissolved ICPP calcines to minimize high level waste volumes and economic penalties perceived for final disposal of these wastes. The actinide separation process being emphasized for the dissolved calcine solutions is the TRUEX process. Substantial problems have been encountered during TRUEX flowsheet development efforts for dissolved zirconium calcine simulant due to the high concentrations and subsequent extraction of zirconium from the feed. Alteration of the calcine dissolution parameters has resulted in the development of a successful TRUEX/Zr calcine baseline flowsheet. This flowsheet has been tested using 22 stages of a 2.0 centimeter diameter centrifugal contactor pilot plant using simulated dissolved Zr calcine solution. With this flowsheet, a removal efficiency of > 96% was obtained for {sup 241}Am (analytical detection limits were reached). Less than 0.25% of the {sup 95}Zr exited with the high-level waste strip product.

Law, J.D.; Herbst, R.S.; Brewer, K.N.; Todd, T.A.

1998-07-01T23:59:59.000Z

398

Microscopic Description of Nuclear Fission: Fission Barrier Heights of Even-Even Actinides  

E-Print Network (OSTI)

We evaluate the performance of modern nuclear energy density functionals for predicting inner and outer fission barrier heights and energies of fission isomers of even-even actinides. For isomer energies and outer barrier heights, we find that the self-consistent theory at the HFB level is capable of providing quantitative agreement with empirical data.

J. McDonnell; N. Schunck; W. Nazarewicz

2013-01-31T23:59:59.000Z

399

Method of loading organic materials with group III plus lanthanide and actinide elements  

DOE Patents (OSTI)

Disclosed is a composition of matter comprising a tributyl phosphate complex of a group 3, lanthanide, actinide, or group 13 salt in an organic carrier and a method of making the complex. These materials are suitable for use in solid or liquid organic scintillators, as in x-ray absorption standards, x-ray fluorescence standards, and neutron detector calibration standards.

Bell, Zane W. (Oak Ridge, TN); Huei-Ho, Chuen (Oak Ridge, TN); Brown, Gilbert M. (Knoxville, TN); Hurlbut, Charles (Sweetwater, TX)

2003-04-08T23:59:59.000Z

400

Microscopic Description of Nuclear Fission: Fission Barrier Heights of Even-Even Actinides  

E-Print Network (OSTI)

We evaluate the performance of modern nuclear energy density functionals for predicting inner and outer fission barrier heights and energies of fission isomers of even-even actinides. For isomer energies and outer barrier heights, we find that the self-consistent theory at the HFB level is capable of providing quantitative agreement with empirical data.

McDonnell, J; Nazarewicz, W

2013-01-01T23:59:59.000Z

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401

Colloidal products and actinide species in leachate from spent nuclear fuel  

Science Conference Proceedings (OSTI)

Two well-characterized types of spent nuclear fuel (ATM-103 and ATM-106) were subjected to unsaturated leach tests with simulated groundwater at 90{degrees}C. The actinides present in the leachate were determined at the end of two successive periods of {approximately}60 days and after an acid strip done at the end of the second period. Both colloidal and soluble actinide species were detected in the leachates which had pHs ranging from 4 to 7. The uranium phases identified in the colloids were schoepite and soddyite. In addition, the actinide release behavior of the two fuels appeared to be different for both the total amount of material released and the relative amount of each isotope released. This paper will focus on the detection and identification of the colloidal species observed in the leachate that was collected after each of the first two successive testing periods of approximately 60 days each. In addition, preliminary values for the total actinide release for these two periods are reported.

Finn, P.A.; Buck, E.C.; Gong, M.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-12-31T23:59:59.000Z

402

Possible experimental signature of octupole correlations in the 0$^+_2$ states of the actinides  

E-Print Network (OSTI)

$J^{\\pi}$= 0$^+$ states have been investigated in the actinide nucleus ${}^{240}$Pu up to an excitation energy of 3 MeV with a high-resolution (p,t) experiment at $E_{p}$= 24 MeV. To test the recently proposed $J^{\\pi}$= 0$^+_2$ double-octupole structure, the phenomenological approach of the spdf-interacting boson model has been chosen. In addition, the total 0$^+$ strength distribution and the $0^+$ strength fragmentation have been compared to the model predictions as well as to the previously studied (p,t) reactions in the actinides. The results suggest that the structure of the 0$^+_2$ states in the actinides might be more complex than the usually discussed pairing isomers. Instead, the octupole degree of freedom might contribute significantly. The signature of two close-lying 0$^+$ states below the 2-quasiparticle energy is presented as a possible manifestation of strong octupole correlations in the structure of the 0$^+_2$ states in the actinides.

M. Spieker; D. Bucurescu; J. Endres; T. Faestermann; R. Hertenberger; S. Pascu; S. Skalacki; S. Weber; H. -F. Wirth; N. -V. Zamfir; A. Zilges

2013-10-21T23:59:59.000Z

403

Application of Feed and Bleed Operations to Remove High Level ...  

Cleaning Method Phase Date. 5 Process Identification • After Mechanical Sludge Removal and Chemical Cleaning: ... Block Diagram Filtrate Solids Separation Solids Slurry

404

Nitrogen removal from natural gas using two types of membranes ...  

A process for treating natural gas or other methane-rich gas to remove excess nitrogen. The invention relies on two-stage membrane separation, using ...

405

Conventional methods for removing sulfur and other contaminants...  

NLE Websites -- All DOE Office Websites (Extended Search)

Conventional methods for removing sulfur and other contaminants from syngas typically rely on chemical or physical absorption processes operating at low temperatures. When cooled...

406

NETL: News Release - Innovative Mercury Removal Technique Shows Early  

NLE Websites -- All DOE Office Websites (Extended Search)

August 5, 2003 August 5, 2003 Innovative Mercury Removal Technique Shows Early Promise Photochemical Process Developed in Federal Lab Removes Mercury from Flue Gas - NETL scientist Evan Granite prepares a lab test of the UV mercury removal process. - NETL scientist Evan Granite prepares for a lab test of the UV mercury removal process. MORGANTOWN, WV - A promising technology to remove mercury from coal-fired power plants -- dubbed the "GP-254 Process" -- has been developed and is currently being tested at the Department of Energy's National Energy Technology Laboratory (NETL). Newly patented, the GP-254 Process enhances mercury removal using ultraviolet light to induce various components of power plant stack gas to react with the mercury, and changes the

407

The contrasting fission potential-energy structure of actinides and mercury isotopes  

E-Print Network (OSTI)

Fission-fragment mass distributions are asymmetric in fission of typical actinide nuclei for nucleon number $A$ in the range $228 \\lnsim A \\lnsim 258$ and proton number $Z$ in the range $90\\lnsim Z \\lnsim 100$. For somewhat lighter systems it has been observed that fission mass distributions are usually symmetric. However, a recent experiment showed that fission of $^{180}$Hg following electron capture on $^{180}$Tl is asymmetric. We calculate potential-energy surfaces for a typical actinide nucleus and for 12 even isotopes in the range $^{178}$Hg--$^{200}$Hg, to investigate the similarities and differences of actinide compared to mercury potential surfaces and to what extent fission-fragment properties, in particular shell structure, relate to the structure of the static potential-energy surfaces. Potential-energy surfaces are calculated in the macroscopic-microscopic approach as functions of fiveshape coordinates for more than five million shapes. The structure of the surfaces are investigated by use of an immersion technique. We determine properties of minima, saddle points, valleys, and ridges between valleys in the 5D shape-coordinate space. Along the mercury isotope chain the barrier heights and the ridge heights and persistence with elongation vary significantly and show no obvious connection to possible fragment shell structure, in contrast to the actinide region, where there is a deep asymmetric valley extending from the saddle point to scission. The mechanism of asymmetric fission must be very different in the lighter proton-rich mercury isotopes compared to the actinide region and is apparently unrelated to fragment shell structure. Isotopes lighter than $^{192}$Hg have the saddle point blocked from a deep symmetric valley by a significant ridge. The ridge vanishes for the heavier Hg isotopes, for which we would expect a qualitatively different asymmetry of the fragments.

Takatoshi Ichikawa; Akira Iwamoto; Peter Möller; Arnold J. Sierk

2012-03-09T23:59:59.000Z

408

THERMALLY SHIELDED MOISTURE REMOVAL DEVICE  

DOE Patents (OSTI)

An apparatus is presented for removing moisture from the air within tanks by condensation upon a cartridge containing liquid air. An insulating shell made in two halves covers the cartridge within the evacuated system. The shell halves are hinged together and are operated by a system of levers from outside the tank with the motion translated through a sylphon bellows to cover and uncover the cartridge. When the condensation of moisture is in process, the insulative shell is moved away from the liquid air cartridge, and during that part of the process when there is no freezing out of moisture, the shell halves are closed on the cell so thnt the accumulated frost is not evaporated. This insulating shell greatly reduces the consumption of liquid air in this condensation process.

Miller, O.E.

1958-08-26T23:59:59.000Z

409