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in subjects: "Transuranium Elements 050800* -- Nuclear Fuels-- Spent Fuels Reprocessing"

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Coprocessing of thermal reactor fuels

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 feasibi… more

Date: January 1, 1985

Creator: Ballard, W.W. Jr.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Mechanism of plutonium metal dissolution in HNO/sub 3/-HF-N/sub 2/H/sub 4/ solution

An oxidation-reduction balance of the products of the dissolution of plutonium metal and alloys in HNO/sub 3/-HF-N/sub 2/H/sub 4/ solution shows that the major reactions during dissolution are the reduction of nitrate to NH/sub 3/, N/sub 2/ and N/sub 2/O by the metal, and the oxidation of H free radicals to NH/sub 3/ by N/sub 2/H/sub 4/. Reactions between HNO/sub 3/ and N/sub 2/H/sub 4/ produce varying amounts of HN/sub 3/. The reaction rate is greater for delta-Pu than alpha-Pu, and is increas… more

Date: January 1, 1985

Creator: Karraker, D G

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Electrorefining of uranium and plutonium from liquid cadmium

Feasibility of electrorefining of U, Pu, and mixtures thereof using a liquid Cd anode and a molten-salt electrolyte was investigated for the proposed pyrometallurgical process for the Integral Fast Reactor fuel. (DLC)

Date: January 1, 1985

Creator: Tomczuk, Z.; Poa, D. S.; Miller, W. E. & Steunenberg, R. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Tests of alternative reductants in the second uranium purification cycle

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Purex process

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Mixing and solid suspension in a stirred precipitator

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 … more

Date: January 1, 1986

Creator: Chang, T P

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Recovery of plutonium from HEPA filters by Ce(IV)-promoted dissolution of PuO/sub 2/ and recycle of the cerium promoter

The experimental studies carried out included (1) the electrolytic production of Ce(IV) from Ce(III), (2) the leaching of refractory PuO/sub 2/ from HEPA filter materials with maintenance of Ce(IV) concentrations by anodic oxidation during leaching, and (3) evaluation of methods for contacting the HEPA solids with the leaching solution and for separating the solid residue from the leaching liquor. Anodic oxidation of Ce(III) was accomplished with an electric current efficiency of about 85% at c… more

Date: January 1, 1980

Creator: Leuze, R. E.; Bond, W. D. & Scheitlin, F. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Further development of IDGS: Isotope dilution gamma-ray spectrometry

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 spik… more

Date: January 1, 1991

Creator: Li, T. K.; Parker, J. L.; Kuno, Y.; Sato, S.; Kamata, M. & Akiyama, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Purex: process and equipment performance

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 th… more

Date: January 1, 1986

Creator: Orth, D.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Studies of the hydrolytic and gamma-radiolytic degradation of the TRUEX-CC1/sub 4/ process solvent

The stability of the TRUEX-CCl/sub 4/ solvent (0.25M OphiD(iB)CMPO-0.75M TBP-CCl/sub 4/) toward hydrolytic and gamma-radiolytic degration is presently under study. The experimental procedure is discussed. Reported this quarter are the complete results for experiments run at 50/sup 0/C to measure the nitric-acid-catalyzed degradation of the TRUEX-CCl/sub 4/ solvent and the partial results for hydrolytic degradation experiments run at 70/sup 0/C and gamma-radiolytic degradation experiments run at… more

Date: May 1, 1985

Creator: Vandegrift, G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Alternate extractants to tributyl phosphate for reactor fuel reprocessing

Both tri(n-hexyl) phosphate (THP) and tri(2-ethylhexyl) phosphate (TEHP) have some important potential process advantages over TBP for reactor fuel reprocessing. These include negligible aqueous phase solubility and less tendency toward third phase and crud formation. The alkyl chain branching of TEHP makes it much more stable to chemical degradation than TBP and probably also accounts for its much weaker ruthenium extraction. The higher uranium and plutonium extraction power of THP and TEHP al… more

Date: January 1, 1983

Creator: Crouse, D. J.; Arnold, W. D. & Hurst, F. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Plutonium-uranium separation in the Purex process using mixtures of hydroxylamine nitrate and ferrous sulfamate

Laboratory studies, followed by plant operation, established that a mixture of hydroxylamine nitrate (HAN) and ferrous sulfamate (FS) is superior to FS used alone as a reductant for plutonium in the Purex first cycle. FS usage has been reduced by about 70% (from 0.12 to 0.04M) compared to the pre-1978 period. This reduced the volume of neutralized waste due to FS by 194 liters/metric ton of uranium (MTU) processed. The new flowsheet also gives lower plutonium losses to waste and at least compar… more

Date: November 1, 1983

Creator: McKibben, J. M.; Chostner, D. F. & Orebaugh, E. G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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

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-catal… more

Date: January 1, 1990

Creator: Horwitz, E.P.; Diamond, H.; Gatrone, R.C.; Nash, K.L. & Rickert, P.G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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

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.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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

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.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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

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, reso… more

Date: January 1, 1988

Creator: Strunk, W.D.; Singh, S.P. & Tuft, R.M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Actinide recovery from pyrochemical residues

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 … more

Date: May 1, 1985

Creator: Avens, Larry R.; Clifton, David G. & Vigil, Alvin R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Corrosion studies in molten calcium chloride with chlorine

This study is aimed at testing new materials for use in molten salt processing of plutonium. Because of the high corrosiveness of chlorine, present materials have a high rate of failure. Materials less subject to corrosion are needed to minimize costs resulting from rapid failure of sparge tubes, stirring apparatus, and crucibles; to reduce the quantity of plutonium-contaminated scrap; and to improve the purity of the plutonium product. The processing environment of molten CaCl{sub 2}--CaO salt… more

Date: January 1, 1990

Creator: McLaughlin, D. F.; Sessions, C. E. & Marra, J. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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

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 (avai… more

Date: January 1, 1984

Creator: Easter, M. E.; Dodds, H. L. & Primm, R. T., III

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Applicability of hydroxylamine nitrate reductant in pulse-column contactors

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. Puls… more

Date: May 1, 1983

Creator: Reif, D.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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

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,… more

Date: December 7, 1990

Creator: Muscatello, A. C. & Killion, M. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Review of major plutonium pyrochemical technology

The past twenty years have seen significant growth in the development and application of pyrochemical technology for processing of plutonium. For particular feedstocks and specific applications, non-aqueous high-temperature processes offer key advantages over conventional hydrometallurgical systems. Major processes in use today include: (1) direct oxide reduction for conversion of PuO/sub 2/ to metal, (2) molten salt extraction for americium removal from plutonium, (3) molten salt electrorefini… more

Date: January 1, 1983

Creator: Moser, W.S. & Navratil, J.D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department