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A CONTINUOUS DAREX PROCESS FLOWSHEET

Description: A Darex flowsheet is presented wherein dissolution of stainiess steel- type fuel elements in dilute aqua regia, stripping of chloride from spent dissolver solution, adjustment of stripper product composition and rectification of HNO/sub 3/ --H/sub 2/O vapors are accomplished in a completely continuous relatively simple loop process. Product from the loop needs only dilution with water to be made suitable as feed for a solvent extraction processing plant. The proposed scheme differs from earlier concepts. A continuous, pot dissolver (well- mixed liquid phase) is utilized rather than a trickle type dissolver. A strip vapor slightly lower in HNO/sub 3/ concentration than the azeotrope is used. A feed adjustment tank serves not only to recover HNO/sbu 3/ --H/sub 2/O vapor, but also to dehydrate any silica present and to adjust the product composition such that upon water addition the material is ready for feed to solvent extraction. In most previous concepts, separate, batchwise feed adjustment was contemplated. (auth)
Date: October 10, 1957
Creator: Kitts, F.G.
Partner: UNT Libraries Government Documents Department

The Use of Boron for Fluoride Complexing in Thorex Dissolver Solutions

Description: Preliminary measurements of the corrosion of titanium were made in 13M HNO/sub 3/ -0.05M fluoride using O.1M H/sub 3/BO/sub 3/ as a liquid and vapor- phase complexing agent. Titanium Ax-55 was attacked at average rates of 0.58 and 0.33 mil/ month in the liquid and vapor. In dissolver solutions containing 0.5 and 1.0M titanium, all rates were less than 0.1 mil/month. (auth)
Date: August 1, 1959
Creator: Kitts, F. G.
Partner: UNT Libraries Government Documents Department

EVALUATION OF AN ENGINEERING DEMONSTRATION OF THE MODIFIED ZIRFLEX AND NEUFLEX PROCESSES FOR THE PREPARATION OF SOLVENT EXTRACTION FEEDS FROM UNIRRADIATED ZIRCONIUM-BASE REACTOR FUELS

Description: In order to recover uranium from zirconium-base reactor fuels by solvent extraction, the metailic fuel and cladding must first be dissolved and a suitable feed solution prepared. Such preparations of solvent extraction feeds were successfully accomplished batchwise using both the Modified Zirflex and Neuflex processes employing an NH/sub 4/F -- oxidant mixture to dissolve the fuel elements, and the feed. (The d Zirflex feed, and H/sub 2/O for the Neuflex feed.) In the Modified Zirflex process, a dissolvent about 6 M in NH/sub 4/F with an excess of H/sub 2/O/sub 2/ to oxidize uranium to the more-soluble U(VI) valence state. The off-gas, after NH/sub 3/ removal, is an H/sub 2/-O/sub 2/ mixture of small volume, which is diluted with air to a safe concentration. Then nitric acid-aluminum nitrate is added to the dissolution product, yielding a solvent extraction feed from which uranium is recovered by using TBP-Amsco as the extractant. In the Neuflex process, the dissolvent is NH/sub 4/F--H/sub 2/O/sub 2/, with less than a stoichiometric amount of NH/sub 4/NO/sub 3/. Without NH/sub 4/NO/sub 3/, the scrubbed off-gas is principally hydrogen, on the hydrogen-rich side of the flammable range of H/sub 2/-O/sub 2/ mixtures, Only water is added to this dissolution product, yielding a neutral fluoride feed from which uranium is extractable by use of Dapex reagents. ln both processes the F: Zr charge ratio, initial surface condition, and maximum section thickness of the fuel element were the principa1 determinants of total dissolution time. The zirconium loading as determined by the free fluoride - zirconium solubility relationship limited the capacity of fuels containing less than 2% U, while the free-fluoride-to-uranium ratio of about 100 required for solution stability was the limiting factor with alloys containing higher percentages of uranium, Hydrogen peroxide concentration was not an important factor in solution stability; ...
Date: March 1, 1964
Creator: Kitts, F.G.
Partner: UNT Libraries Government Documents Department

A PRELIMINARY STUDY OF PRE-SOLVENT EXTRACTION TREATMENT OF STAINLESS STEEL- URANIUM FUELS WITH DILUTE AQUA REGIA

Description: The continuous dissolution of 304 stainless steel and stainless steel - UO/sub 2/ alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver prcduct to remove chloride ion was studied in a 4-in. diameter Pyrek bubblecap column containing 12 single babble cap plates. Continuous dissolution rates and dissolver product stainiess steel loadings were correlatsd with aqua regla feed composition, acid feed rats and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained for various vapor to liquid mole ratios and for several nitric acid stripping vapor concentrations. Noncondensable off-gas compositions and rates were also measured. (auth)
Date: October 11, 1957
Creator: Kitts, F.G. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

Pilot-scale demonstration of the modified direct denitration process to prepare uranium oxide for fuel fabrication evaluation

Description: The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF{sub 6} product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO{sub 3} to be shipped to fabricators for making UO{sub 2} pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO{sub 3} suitable for making reactor-grade fuel pellets.
Date: April 1, 1994
Creator: Kitts, F. G.
Partner: UNT Libraries Government Documents Department

DAREX PROCESSING OF APPR FUEL: EFFECT OF ACIDITY AND GAS SPARGING ON RATE OF CHLORIDE REMOVAL FROM DISSOLVER PRODUCT DURING REFLUXING

Description: The rate of chloride removal varied directly with HNO/sub 3/ concentration fn an APPR-type Darex dissolver product containing 100 g/liter metal loading, 0.58 M initial chloride, and initial HNO/sub 3/ concentrations of 8, 9, 10, 12, and 14 M. The removal rate with 8 and 9 M HNO/sub 3/ was very low. After 6 hr refluxing, the chloride content decreased to 0.50 and 0.36 M, respectively. After refluxing for the same time with 10 to 14 M HNO/sub 3/, the product contained 0.064to 0.0007 M (2270 to 25 ppm) chloride. The effect of air sparging was approximately equivalent to refluxing without sparging at a HNO/sub 3/ concentration 2 M higher. After 6 hr sparging and refluxing the chloride content varied from 0.034 to < 0.00014 M (1200 to < 5 ppm) for initial HNO/sub 3/ concentrations from 8 to 14 M. (auth)
Date: August 1, 1959
Creator: Finney, B.C. & Kitts, F.G.
Partner: UNT Libraries Government Documents Department

Routine and post-accident sampling in nuclear reactors

Description: Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experience.
Date: January 1, 1980
Creator: Armento, W.J.; Kitts, F.G. & German, G.E.
Partner: UNT Libraries Government Documents Department

DAREX PROCESS: PROCESSING OF STAINLESS STEEL-CONTAINING REACTOR FUELS WITH DILUTE AQUA REGIA

Description: The Darex process developed for ihe recovery of U from stainless steel- containing reactor fuels consists of dissolution of the fuel material in dilute aqua regia, removal of chloride from the solution to prevent corrosion of downstream stainless steel process equipment, and adjustment of the nitrate solution to solvent extraction feed conditions. Each step can be either continuous, semi-continuous, or batch with continuous operation showing much higher throughput for comparable equipment. The preferred dissolvent is 5 M HNO/ sub 3/-2 M HCl, since dissolution rates and metal loadings are near maximum. Nitric acid from 60 to 95 wt% can be used in decreasing ihe chloride concentration to <350 ppm; ihe higher strength acids have process advantages. Excess nitric acid is recovered and recycled during produciion of a concentrated metal-salt solution, which is diluted io Purex solvent extraction feed acidity, 2- 3 M HNO/sub 3/. Titanium is a satisfactory material of construction, wiih corrosion rates <l mil/mo in all process environments and over-all heat transfer coefficients comparable to those of stainless steel. (auth)
Date: June 1, 1962
Creator: Kitts, F.G. & Clark, W.E.
Partner: UNT Libraries Government Documents Department

Effect of water in salt repositories. Final report

Description: Additional results confirm that during most of the consolidation of polycrystalline salt in brine, the previously proposed rate expression applies. The final consolidation, however, proceeds at a lower rate than predicted. The presence of clay hastens the consolidation process but does not greatly affect the previously observed relationship between permeability and void fraction. Studies of the migration of brine within polycrystalline salt specimens under stress indicate that the principal effect is the exclusion of brine as a result of consolidation, a process that evidently can proceed to completion. No clear effect of a temperature gradient could be identified. A previously reported linear increase with time of the reciprocal permeability of salt-crystal interfaces to brine was confirmed, though the rate of increase appears more nearly proportional to the product of sigma ..delta..P rather than sigma ..delta..P/sup 2/ (sigma is the uniaxial stress normal to the interface and ..delta..P is the hydraulic pressure drop). The new results suggest that a limiting permeability may be reached. A model for the permeability of salt-crystal interfaces to brine is developed that is reasonably consistent with the present results and may be used to predict the permeability of bedded salt. More measurements are needed, however, to choose between two limiting forms of the model.
Date: September 1, 1983
Creator: Baes, C.F. Jr.; Gilpatrick, L.O.; Kitts, F.G.; Bronstein, H.R. & Shor, A.J.
Partner: UNT Libraries Government Documents Department

Darex Process: Processing of Stainless Steel-Containing Reactor Fuels with Dilute Aqua Regia

Description: From abstract: "The Darex process developed for the recovery of uranium from stainless steel-containing reactor fuels consists of three steps: (1) dissolution of the fuel material in dilute aqua regia, (2) removal of chloride from the solution to prevent corrosion of downstream stainless steel process equipment, and (3) adjustment of the nitrate solution to solvent extraction feed conditions."
Date: June 7, 1962
Creator: Kitts, F. G.
Partner: UNT Libraries Government Documents Department

Fuel cycles using adulterated plutonium

Description: Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with /sup 238/U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial.
Date: January 1, 1978
Creator: Brooksbank, R.E.; Bigelow, J.E.; Campbell, D.O.; Kitts, F.G. & Lindauer, R.B.
Partner: UNT Libraries Government Documents Department

Operating experience with a near-real-time inventory balance in a nuclear-fuel-cycle plant

Description: The principal objective of the ORNL Integrated Safeguards Program (ISP) is to provide enhanced material accountability, improved process control, and greater security for nuclear fuel cycle facilities. With the improved instrumentation and computer interfacing currently installed, the ORNL /sup 233/U Pilot Plant has demonstrated capability of a near-real-time liquid-volume balance in both the solvent-extraction and ion-exchange systems. Future developments should include the near-real-time mass balancing of special nuclear materials as both a static, in-tank summation and a dynamic, in-line determination. In addition, the aspects of site security and physical protection can be incorporated into the computer monitoring.
Date: January 1, 1981
Creator: Armento, W.J.; Box, W.D.; Kitts, F.G.; Krichinsky, A.M.; Morrison, G.W. & Pike, D.H.
Partner: UNT Libraries Government Documents Department

Operating experience with a near-real-time inventory balance in a nuclear fuel cycle plant

Description: The principal objective of the ORNL Integrated Safeguards Program (ISP) is to provide enhanced material accountability, improved process control, and greater security for nuclear fuel cycle facilities. With the improved instrumentation and computer interfacing currently installed, the ORNL /sup 233/U Pilot Plant has demonstrated capability of a near-real-time liquid-volume balance in both the solvent-extraction and ion-exchange systems. Future developments should include the near-real-time mass balancing of special nuclear materials as both a static, in-tank summation and a dynamic, in-line determination. In addition, the aspects of site security and physical protection can be incorporated into the computer monitoring.
Date: January 1, 1981
Creator: Armento, W.J.; Box, W.D.; Kitts, F.G.; Krichinsky, A.M.; Morrison, G.W. & Pike, D.H.
Partner: UNT Libraries Government Documents Department

Implementation of the engineering safeguards program (ESP) into nuclear fuel recycle facilities

Description: The principal objective of ORNL-ESP is to demonstrate process monitoring as it might be accomplished by inspectors of any nuclear fuel recycle facility. Improved instrumentation and computer interfacing, currently being installed, provide the ORNL /sup 233/U Pilot Plant with the capability of a dynamic volume balance in the solvent extraction system. Later, an accurate, (almost) instantaneous fissile mass balance will be routinely obtainable in the Pilot Plant. Subsidiary objectives include minimizing MUF/LEMUF, detecting material diversions, and alerting appropriate authorities in control of the facility in case of process anomalies. A continuing program will examine technology which might be utilized for facility design. Ultimately, process monitoring/control integrated with safeguards can convert the ORNL /sup 233/U Pilot Plant into a partial safeguards demonstration facility.
Date: 1979-07~
Creator: Armento, W. J.; Box, W. D.; Brooksbank, R. E.; Kitts, F. G.; Krichinsky, A. M. & Parrott, J. R., Sr.
Partner: UNT Libraries Government Documents Department

Solvent extraction studies with high-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

Description: A batch of high-burnup fuel from the Fast Flux Test Facility (FFTF) was processed in the Solvent Extraction Test Facility (SETF) during Campaign 9. The fuel had a burnup of {similar_to}0 MWd/kg and a cooling time of {similar_to} year. Two runs were made with this fuel; in the first, the solvent contained 30% tri-n-butyl phosphate (TBP) and partitioning of the uranium and plutonium was effected by reducing the plutonium with hydroxylamine nitrate (HAN); in the second, the solvent contained 10% TBP and a low operating temperature was used in an attempt to partition without reducing the plutonium valence. The plutonium reoxidation problem, which was present in previous runs that used HAN, may have been solved by lowering the temperature and acidity in the partition contactor. An automatic control system was used to maintain high loadings of heavy metals in the coextraction-coscrub contactor in order to increase its efficiency while maintaining low losses of uranium and plutonium to the aqueous raffinate. An in-line photometer system was used to measure the plutonium concentration in an intermediate extraction stage; and based on this data, a computer algorithm determined the appropriate adjustments in the addition rate of the extractant. The control system was successfully demonstrated in a preliminary run with purified uranium. However, a variety of equipment and start up problems prevented an extended demonstration from being accomplished during the runs with the FFTF fuel.
Date: October 1, 1986
Creator: Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G. et al.
Partner: UNT Libraries Government Documents Department

Solvent extraction studies with intermediate-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

Description: In Campaign 8, two batches of irradiated fuel from the Fast Flux Test Facility (FFTF) were processed, using 30% TBP-NPH, in the Solvent Extraction Test Facility (SETF). The burnups were about 36 and 55 MWd/kg with 1.3- and 1-year cooling times, respectively. The latter fuel had the highest burnup and shortest cooling time of any fuel ever handled in the SETF. No major problems were noted during the operation of the mixer-settlers, and low uranium and plutonium losses (&lt;0.02%) were achieved. Zirconium and ruthenium decontamination factors (DFs) were improved by increasing the number of scrub stages and increasing the peak solvent loading in the coextraction-coscrub bank. The use of an in-line photometer to measure the uranium and plutonium concentrations in a process stream permitted high solvent loadings of heavy metals to be achieved in the extraction bank while maintaining low losses to the aqueous raffinate. The investigation of two flowsheet options for making separate uranium and plutonium products (organic backscrub and selective uranium extraction) that was started in Campaign 7 was continued. High-quality products were again obtained (uranium and plutonium DFs of {similar_to}0{sup 4}). Plutonium reoxidation was still extensive even though hydrazine was added to the aqueous strip for the organic backscrub flowsheet. Two different plutonium oxalate precipitation procedures [Pu(III) and Pu(IV)] were used in the preparation of the plutonium oxide products; this was done so that the fuel fabrication characteristics of the oxide from the two procedures could be compared. A total of {similar_to}50 g of plutonium was recovered and shipped to the fuel refabrication program.
Date: April 1, 1986
Creator: Benker, D. E.; Bigelow, J. E.; Bond, W. D.; Chattin, F. R.; King, L. J.; Kitts, F. G. et al.
Partner: UNT Libraries Government Documents Department

Solvent extraction studies with low-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

Description: A batch of irradiated Fast Flux Test Facility (FFTF) fuel was processed for the first time in the Solvent Extraction Test Facility (SETF) at the Oak Ridge National Laboratory (ORNL) during Campaign 7. The average burnup of the fuel was only 0.2 atom %, but the cooling time was short enough ({similar_to}2 years) so that {sup 95}Zr was detected in the feed. This short cooling permitted our first measurement of {sup 95}Zr decontamination factors (DFs) without having to use tracers. No operational problems were noted in the operation of the extraction-scrubbing contactor, and low uranium and plutonium losses (< 0.01%) were measured. Fission product DFs were improved noticeably by increasing the number of scrub stages from six to eight. Two flowsheet options for making pure uranium and plutonium products (total partitioning) were tested. Each flowsheet used hydroxylamine nitrate for reducing plutonium. Good products were obtained (uranium DFs of > 10{sup 3} and plutonium DFs of > 10{sup 4}), but each flowsheet was troubled with plutonium reoxidation. Adding hydrazine and lowering the plutonium concentration lessened the problem but did not eliminate it. About 370 g of plutonium was recovered from these tests, purified by anion exchange, converted to PuO{sub 2}, and transferred to the fuel refabrication program. 7 references.
Date: January 1, 1985
Creator: Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G. et al.
Partner: UNT Libraries Government Documents Department

Solvent extraction flowsheet studies with irradiated fuel from the Fast Flux Test Facility

Description: A small batch (approx. 2 kg) of irradiated fuel from the Fast Flux Test Facility (FFTF) was recently used as feed for some solvent extraction experiments that were made in the Solvent Extraction Test Facility (SETF). This fuel, which was from series DEA-1, had been irradiated to a burnup of approx. 0.26 TJ/kg (approx. 3000 MWd/t) and cooled for approx. 1 year. The SETF is located in one of the heavily shielded hot cells of the Transuranium Processing Plant at the Oak Ridge National Laboratory. This facility has been used during the past several years to demonstrate the suitability of conceptual flowsheets for the solvent extraction processing of irradiated light-water reactor and fast breeder reactor fuels. Results of these experiments have provided information on uranium and plutonium separation and recoveries, fission product decontamination, comparisons of flowsheet options, and general operability of the system.
Date: January 1, 1984
Creator: Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Cagle, E.B.; Chattin, F.R.; King, L.J. et al.
Partner: UNT Libraries Government Documents Department

Solvent extraction studies of 10% TBP flowsheets in the solvent extraction test facility using irradiated fuel from the Fast Flux Test Facility

Description: Two solvent extraction experiments were made in the Solvent Extraction Test Facility (SETF) during Campaign 10 to continue the evaluation of: (1) a computer control system for the coextraction-coscrub contractor; and (2) a partitioning technique that separates uranium and plutonium without the aid of chemical reductants. The Fast Flux Test Facility (FFTF) fuel used in this campaign had burnups of {approximately}55 and {approximately}60 (average) MWd/kg. During both experiments, the computer control system successfully maintained stable, efficient operation. The control system used an in-line photometer to monitor the plutonium concentration in the extraction section; and based on this data, it adjusted the addition rate of the extractant to maintain high loadings of heavy metal in the solvent and low raffinate losses. The uranium and plutonium partitioning relied entirely on the differences between the U(VI) and Pu(IV) distribution coefficients (since no reductant was used to adjust the plutonium valence). In order to enhance this difference, the TBP concentration and operating temperature were relatively low in comparison to traditional Purex flowsheets. Final product purities of 99{percent} were achieved for both the uranium and plutonium in one cycle of partitioning.
Date: March 1, 1988
Creator: Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Campbell, D.O.; Chattin, F.R.; King, L.J. et al.
Partner: UNT Libraries Government Documents Department