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Mixing and solid suspension in a stirred precipitator

Description: Full-scale mixing and solid suspension studies have been conducted to determine the optimum agitator design for precipitators used in plutonium processing. Design considerations include the geometry of precipitator vessels, feed locations, flow patterns, and product requirements. Evaluations of various agitator designs are based on their capabilities: (1) to achieve uniform mixing of reactants in minimum time, (2) to suspend the slurry uniformly throughout the vessel, and (3) to minimize power consumption without inducing air entrainment. Tests of full-scale agitator designs showed that significant improvements in mixing, solid suspension, and energy consumption were achieved.
Date: January 1, 1986
Creator: Chang, T P
Partner: UNT Libraries Government Documents Department

Recent development in pyrochemistry at Los Alamos

Description: Recent developments in pyrochemical processing at Los Alamos include the recovery of plutonium from anodes and impure metal by pyroredox and new molten salt handling and purification techniques. The anode is dissolved in a ZnCl/sub 2/ KCl salt to form PuCl/sub 3/ and a zinc and impurities button. Calcium reduction of the PuCl/sub 3/ yields 95 to 98% pure plutonium. New techniques for transferring molten salt from a purification or regeneration vessel to molds has been successfully developed and demonstrated. Additional salt work involving recycle of direct oxide reduction salts using anhydrous hydrogen chloride, phosgene, and chlorine gases is under way. 13 figures, 1 table.
Date: January 1, 1984
Creator: McNeese, J.A.; Fife, K.W. & Williams, J.D.
Partner: UNT Libraries Government Documents Department

Evaluation of nonaqueous processes for nuclear materials

Description: A working group was assigned the task of evaluating the status of nonaqueous processes for nuclear materials and the prospects for successful deployment of these technologies in the future. In the initial evaluation, the study was narrowed to the pyrochemical/pyrometallurgical processes closely related to the processes used for purification of plutonium and its conversion to metal. The status of the chemistry and process hardware were reviewed and the development needs in both chemistry and process equipment technology were evaluated. Finally, the requirements were established for successful deployment of this technology. The status of the technology was evaluated along three lines: (1) first the current applications were examined for completeness, (2) an attempt was made to construct closed-cycle flow sheets for several proposed applications, (3) and finally the status of technical development and future development needs for general applications were reviewed. By using these three evaluations, three different perspectives were constructed that together present a clear picture of how complete the technical development of these processes are.
Date: December 1, 1983
Creator: Musgrave, B.C.; Grens, J.Z.; Knighton, J.B. & Coops, M.S.
Partner: UNT Libraries Government Documents Department

Purex process

Description: The following aspects of the Purex Process are discussed: head end dissolution, first solvent extraction cycle, second plutonium solvent extraction cycle, second uranium solvent extraction cycle, solvent recovery systems, primary recovery column for high activity waste, low activity waste, laboratory waste evaporation, vessel vent system, airflow and filtration, acid recovery unit, fume recovery, and discharges to seepage basin. (LK)
Date: January 1, 1977
Creator: Starks, J.B.
Partner: UNT Libraries Government Documents Department

Calculated k-effectives for plutonium critical experiments. Consolidated Fuel Reprocessing Program

Description: Design criteria for a reprocessing facility for Liquid Metal Fast Breeder Reactor fuel are presently being developed. One major issue of concern is the criticality safety of all equipment (dissolver, centrifuge, holding tanks, etc.) that is used to contain the plutonium solution. The purpose of this work is to evaluate the validity of the SCALE code system for application to plutonium systems when used with cross section data from the 27-group ENDF/B-IV and 16-group Hansen-Roach libraries (available in SCALE). Previous work has been done in this area, but it was limited to one-dimensional discrete ordinates calculations. Twelve sets of critical plutonium experiments yielding a total of thirty-eight computational models are considered. The experiments were performed in spherical, cylindrical, and slab geometries covering a wide range of fuel composition. The hydrogen-to-plutonium atom ratio varied from 3695 for an infinite dilute system to 0.04 for damp, plutonium oxide polystyrene experiments. The SCALE system employs several codes for criticality calculations, but only two are used in this work. Criticality Safety Analysis Sequence Number 1 (CSAS1) uses discrete ordinates theory to calculate k-eff, and its application is limited to one-dimensional calculations. Criticality Safety Analysis Sequence Number 2 (CSAS2) uses Monte Carlo methods to calculate k-eff, and its application includes three dimensional systems. Calculated results of the 38 models using both cross sections libraries are presented. The calculated k-effectives vary from a high of 1.04119 +- 4.87E-3 to a low of 0.99458 +- 4.66E-3. The majority of the k-effectives vary between 1.01 and 1.03 illustrating a systematic overestimation of k-eff. Some of the one dimensional experiments were modeled with both CSAS1 and CSAS2 to see if the calculated k-effectives show any significant dependency on the code used.
Date: January 1, 1984
Creator: Easter, M.E.; Dodds, H.L. & Primm, R.T. III
Partner: UNT Libraries Government Documents Department

Actinide recovery from pyrochemical residues

Description: We demonstrated a new process for recovering plutonium and americium from pyrochemical waste. The method is based on chloride solution anion exchange at low acidity, or acidity that eliminates corrosive HCl fumes. Developmental experiments of the process flow chart concentrated on molten salt extraction (MSE) residues and gave >95% plutonium and >90% americium recovery. The recovered plutonium contained <500 ppM americium and <2500 ppM magnesium. The process operates by sorbing PuCl/sub 6//sup 2 -/ from high-chloride low-acid solution. Americium and other metals are washed from the ion exchange column with lN HNO/sub 3/-4.8M NaCl. After elution, plutonium is recovered by hydroxide precipitation, and americium is recovered by NaHCO/sub 3/ precipitation. All filtrates from the process can be discardable as low-level contaminated waste. Production-scale experiments are in progress for MSE residues. Flow charts for actinide recovery from electro-refining and direct oxide reduction residues are presented and discussed.
Date: May 1, 1985
Creator: Avens, L.R.; Clifton, D.G. & Vigil, A.R.
Partner: UNT Libraries Government Documents Department

The effect of fluoride and aluminum on the anion exchange of plutonium from nitric acid

Description: Anion exchange in nitric acid is a prominent aqueous process used to recover and purify plutonium from impure nuclear materials. This process is sensitive to fluoride ion, which complexes plutonium in competition with the anionic nitrato complex that is strongly sorbed on the anion exchange column. Fluoride interference traditionally has been counteracted by adding a masking agent, such as aluminum, that forms competing complexes with fluoride. The interfering effect of fluoride is known to be a function not only of the fluoride-to-aluminum ratio but also of the fluoride-to-plutonium ratio. This report summarizes a Los Alamos study of the effect of 25 fluoride-aluminum-plutonium conmbinations on the anion exchange sorption of plutonium. Five aluminum-to-plutonium ratios ranging from 0.10 to 10 were each evaluated at five fluoride-to-aluminum ratios that ranged from 0 to 6. The fluoride-to-plutonium ratio has a greater influence on plutonium sorption than does the fluoride-to-aluminum ratio. Aluminum was less effective as a masking agent than had been assumed, because measurable fluoride interference occurred at all levels of added aluminum.
Date: July 1, 1987
Creator: Marsh, S.F.
Partner: UNT Libraries Government Documents Department

Plutonium and americium processing chemistry and technology

Description: Plutonium processing in the USA originated at Hanford and Los Alamos as part of the Manhattan Project in 1943. Hanford separated plutonium from irradiated reactor fuel, whereas Los Alamos purified the plutonium, as well as recovered plutonium from residues and scrap. In the early 1950's, similar processing facilities were constructed at Savannah River and Rocky Flats. The PUREX process (tri-n-butyl phosphate extraction) is used at Hanford and Savannah River plants to separate plutonium from irradiated reactor fuel. At Los Alamos and the Rocky Flats Plant (RFP), both pyrochemical and aqueous processes are used to recover and purify plutonium. A by-product in the plutonium recovery processes is americium-241 from the beta decay of plutonium-241 present in the plutonium-239 stream. An overview of the americium and plutonium processing chemistry and technology at RFP is presented. 49 references, 3 figures.
Date: January 1, 1984
Creator: Navratil, J.D.
Partner: UNT Libraries Government Documents Department

Acid-split flowsheets for uranium-plutonium partitioning without a reductant

Description: The flowsheet discussed has been tested in a hot cell experiment using 10% TBP and a poorly controlled temperature near 15/sup 0/C. The test was carried out in the Solvent Extraction Test Facility at Oak Ridge National Laboratory, using highly irradiated mixed-oxide fuel from the Fast Flux Test Facility reactor at Hanford, Washington. The observed concentration profiles for U, Pu, and acid are shown graphically.
Date: January 1, 1986
Creator: Campbell, D.O.; Crouse, D.J. & Mills, A.L.
Partner: UNT Libraries Government Documents Department

Evaluation of anion exchange resins for processing plutonium--neptunium residues

Description: An anion exchange process was developed to process miscellaneous residues of plutonium plus 0.5 wt % neptunium to allow prompt return of the plutonium to a plutonium recovery process. Several macroreticular anion exchange resins were compared to Dowex 1-X4 for the process. Dowex 1-X4 showed the best performance for the plutonium (III)-neptunium(IV) separation.
Date: August 20, 1977
Creator: Navratil, J. D. & Leebl, R. G.
Partner: UNT Libraries Government Documents Department

Pyrochemical processing of plutonium. Technology review report

Description: Non-aqueous processes are now in routine use for direct conversion of plutonium oxide to metal, molten salt extraction of americium, and purification of impure metals by electrorefining. These processes are carried out at elevated temperatures in either refractory metal crucibles or magnesium-oxide ceramics in batch-mode operation. Direct oxide reduction is performed in units up to 700 gram PuO/sub 2/ batch size with molten calcium metal as the reductant and calcium chloride as the reaction flux. Americium metal is removed from plutonium metal by salt extraction with molten magnesium chloride. Electrorefining is used to isolate impurities from molten plutonium by molten salt ion transport in a controlled potential oxidation-reduction cell. Such cells can purify five or more kilograms of impure metal per 5-day electrorefining cycle. The product metal obtained is typically > 99.9% pure, starting from impure feeds. Metal scrap and crucible skulls are recovered by hydriding of the metallic residues and recovered either as impure metal or oxide feeds.
Date: September 8, 1982
Creator: Coops, M.S.; Knighton, J.B. & Mullins, L.J.
Partner: UNT Libraries Government Documents Department

Coprocessing of thermal reactor fuels

Description: The Nuclear Power Development Division (NPD) under the Assistant Secretary for Energy Technology in the Department of Energy (DOE) is responsible for examining alternative nuclear reactor fuel recycle systems which have potential for reducing the risk of proliferation of nuclear weapons. NPD is administering a base technology program of research and development and design studies which will provide a sound technical foundation for evaluating the nonproliferation potential and commercial feasibility of these alternatives. The Savannah River Laboratory (SRL) has been assigned as the technical lead for those activities associated with the processing of thermal reactor fuel. In order to systematically identify technical requirements and design solutions, SRL periodically updates a Design Integration Study (DIS). The reference process being incorporated into the current DIS is coprocessing uranium and plutonium in a manner whereby pure plutonium is never available in a separate stream. As with other processes, coprocessing doesn't offer a technical fix for preventing proliferators. A flowsheet for this reference process is described with particular emphasis on technical issues and proliferation resistance advantages of coprocessing over conventional Purex processing.
Date: January 1, 1985
Creator: Ballard, W.W. Jr.
Partner: UNT Libraries Government Documents Department

Applicability of hydroxylamine nitrate reductant in pulse-column contactors

Description: Uranium and plutonium separations were made from simulated breeder reactor spent fuel dissolver solution with laboratory-sized pulse column contactors. Hydroxylamine nitrate (HAN) was used for reduction of plutonium (1V). An integrated extraction-partition system, simulating a breeder fuel reprocessing flowsheet, carried out a partial partition of uranium and plutonium in the second contactor. Tests have shown that acceptable coprocessing can be ontained using HAN as a plutonium reductant. Pulse column performance was stable even though gaseous HAN oxidation products were present in the column. Gas evolution rates up to 0.27 cfm/ft/sup 2/ of column cross section were tested and found acceptable.
Date: May 1, 1983
Creator: Reif, D.J.
Partner: UNT Libraries Government Documents Department

Purex: process and equipment performance

Description: The Purex process is the solvent extraction system that uses tributyl phosphate as the extractant for separating uranium and plutonium from irradiated reactor fuels. Since the first flowsheet was proposed at Oak Ridge National Laboratory in 1950, the process has endured for over 30 years with only minor modifications. The spread of the technology was rapid, and worldwide use or research on Purex-type processes was reported by the time of the 1955 Geneva Conference. The overall performance of the process has been so good that there are no serious contenders for replacing it soon. This paper presents: process description; equipment performance (mixer-settlers, pulse columns, rapid contactors); fission product decontamination; solvent effects (solvent degradation products); and partitioning of uranium and plutonium.
Date: January 1, 1986
Creator: Orth, D.A.
Partner: UNT Libraries Government Documents Department

Tests of alternative reductants in the second uranium purification cycle

Description: Miniature mixer-settler tests of the second uranium purification cycle show that plutonium cannot be removed by hydroxylamine-hydrazine (NH/sub 2/OH-N/sub 2/H/sub 4/) because the acidity is too high, or by 2,5-di-t-pentylhydroquinone because HNO/sub 3/ oxidizes the hydroquinone. Plutonium can be removed satisfactorily when U(IV)-hydrazine is used as the reductant.
Date: May 1, 1980
Creator: Thompson, M.C.
Partner: UNT Libraries Government Documents Department

Dynamic considerations in the development of centrifugal separators used for reprocessing nuclear fuel

Description: The development of centrifugal separators has been a key ingredient in improving the process used for reprocessing of spent nuclear fuel. The separators are used to segregate uranium and plutonium from the fission products produced by a controlled nuclear reaction. The separators are small variable speed centrifuges, designed to operate in a harsh environment. Dynamic problems were detected by vibration analysis and resolved using modal analysis and trending. Problems with critical speeds, resonances in the base, balancing, weak components, precision manufacturing, and short life have been solved.
Date: January 1, 1988
Creator: Strunk, W.D.; Singh, S.P. & Tuft, R.M.
Partner: UNT Libraries Government Documents Department

Further development of IDGS: Isotope dilution gamma-ray spectrometry

Description: The isotope dilution gamma-ray spectrometry (IDGS) technique for determining the plutonium concentration and isotopic composition of highly radioactive spent-fuel dissolver solutions has been further developed. Both the sample preparation and the analysis have been improved. The plutonium isotopic analysis is based on high-resolution, low-energy gamma-ray spectrometry. The plutonium concentration in the dissolver solutions then is calculated from the measured isotopic differences among the spike, the dissolver solution, and the spiked dissolver solution. Plutonium concentrations and isotopic compositions of dissolver solutions analyzed from this study agree well with those obtained by traditional isotope dilution mass spectrometry (IDMS) and are consistent with the first IDGS experimental result. With the current detector efficiency, sample size, and a 100-min count time, the estimated precision is {approximately}0.5% for {sup 239}Pu and {sup 240}Pu isotopic analyses and {approximately}1% for the plutonium concentration analysis. 5 refs., 2 figs., 7 tabs.
Date: January 1, 1991
Creator: Li, T.K.; Parker, J.L. (Los Alamos National Lab., NM (United States)); Kuno, Y.; Sato, S.; Kamata, M. & Akiyama, T. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))
Partner: UNT Libraries Government Documents Department

Consolidated fuel reprocessing program. Progress report, October 1--December 31, 1978

Description: The status of the following studies is reported: plutonium (IV) polymer reaction in aqueous solutions; plutonium reductive stripping studies; plutonium--uranium--thorium coprocessing studies; plutonium losses due to solution instability and solids formation; solvent cleanup; nitrogen compound chemistry; fission product chemistry; electrochemical methods evaluation; evaluation of alternate extractants; hot-cell development; solvent extraction; product conversion; analytical chemistry development; voloxidation; dissolution; feed preparation; off-gas processing; and engineering systems. (LK)
Date: March 1, 1979
Creator: Burch, W.D.; Feldman, M.J.; Groenier, W.S.; Vondra, B.L. & Unger, W.E.
Partner: UNT Libraries Government Documents Department

Evaluation of a photo-electron rejecting alpha liquid scintillation (PERALS) spectrometer for the measurement of alpha-emitting radionuclides

Description: Results from the evaluation of a PERALS spectrometer for alpha particle measurements by liquid scintillation counting in samples from the nuclear fuel cycle are presented. Examples of PERALS spectra of process, waste, and environmental samples containing Th, U, Pu and Am from the Savannah River Site are shown. The advantages, disadvantages, and limitations of the PERALS technique are discussed. 15 refs., 11 figs.
Date: January 1, 1990
Creator: Cadieux, J.R.
Partner: UNT Libraries Government Documents Department

Aqueous recovery of actinides from aluminum alloys

Description: Early in the 1980's, a joint Rocky Flats/Savannah River program was established to recover actinides from scraps and residues generated during Rocky Flats purification operations. The initial program involved pyrochemical treatment of Molten Salt Extraction (MSE) chloride salts and Electrorefining (ER) anode heel metal to form aluminum alloys suitable for aqueous processing at Savannah River. Recently Rocky Flats has expressed interest in expanding the aluminum alloy program to include treatment of chloride salt residues from a modified Molten Salt Extraction process and from the Electrorefining purification operations. Samples of the current aluminum alloy buttons were prepared at Rocky Flats and sent to Savannah River Laboratory for flowsheet development and characterization of the alloys. A summary of the scrub alloy-anode heel alloy program will be presented along with recent results from aqueous dissolution studies of the new aluminum alloys. 2 figs., 4 tabs.
Date: January 1, 1989
Creator: Gray, J.H.; Chostner, D.F. & Gray, L.W.
Partner: UNT Libraries Government Documents Department

Results of the metallographic examination of the Ta crucible used in the M. S. E. runs

Description: A cross section from a Ta crucible used in numerous Molten Salt Extraction (MSE) runs was submitted to metallography to determine the soundness of the crucible wall, type of Pu attack, depth of wall penetration by the Pu and general microstructure. The crucible contained molten Pu and Am, with CaCl{sub 2}, KCl and PuCl{sub 3} salts ran at temperatures of 750{degree}C to 900{degree}C for approximately 10 to 12 hours. This report documents the findings of this study.
Date: October 22, 1990
Creator: Furr, J.S.
Partner: UNT Libraries Government Documents Department

TUCS: A new class of aqueous complexing agents for use in solvent extraction processes

Description: The 1-hydroxyethyl-1,1-diphosphonic (HEDPA) and vinylidene-1,1-disphosphonic (VDPA) acids have been studied as stripping agents for Am, Pu, and U from TRUEX process solvent (0.2 M CMPO-1.2 M TBP-dodecane). Both disphosphonic acids have been shown to be highly effective stripping agents using pristine process solvent as well as radiolytically degraded solvent. The actinide-HEDPA and -VDPA complexes are more soluble than the corresponding oxalate complexes and are readily destroyed by metal-catalyzed H{sub 2}O{sub 2}. 12 refs., 4 figs., 2 tabs.
Date: January 1, 1990
Creator: Horwitz, E.P.; Diamond, H.; Gatrone, R.C.; Nash, K.L. & Rickert, P.G.
Partner: UNT Libraries Government Documents Department

Chloride anion exchange coprocessing for recovery of plutonium from pyrochemical residues and Cs sub 2 PuCl sub 6 filtrate

Description: Continuing studies of plutonium recovery from direct oxide reduction (DOR) and electrorefining (ER) pyrochemical process residues show that chloride anion exchange coprocessing is useful and effective. Coprocessing utilizes DOR residue salt as a reagent to supply the bulk of chloride ion needed for the chloride anion exchange process and to improve ER residue salt solubility. ER residue salt and ER scrapeout can be successfully treated, either alone or together, using coprocessing. In addition, chloride anion exchange at 2.0M acidity results in improved process performance by greatly reducing disproportionation of plutonium(IV), eliminating restrictions on oxidation time compared to operation at 1.0M acidity. Laboratory-scale experiments show that below-discard effluent plutonium losses are obtained. Resin capacity was 30 g Pu/{ell} or greater. Furthermore, it is feasible to perform chloride anion exchange recovery of plutonium from filtrate resulting from precipitation of dicesium hexachloroplutonate (Cs{sub 2}PuCl{sub 6}, an oxidant salt to be used in the molten salt extraction process) and integration of its preparation with recovery of DOR salts. 10 refs., 9 figs., 10 tabs.
Date: December 7, 1990
Creator: Muscatello, A.C. & Killion, M.E.
Partner: UNT Libraries Government Documents Department