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Download or read book Dissolution of Irradiated U-Mo Alloy Monolithic Fuel Containing 10% (w/w) Mo written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : Jason L. Schulthess
Publisher :
ISBN 13 : 9781085583084
Total Pages : 174 pages
Book Rating : 4.5/5 (83 download)
Download or read book Degradation of Mechanical Properties of U-Mo Alloy from the Un-irradiated to Irradiated State written by Jason L. Schulthess and published by . This book was released on 2018 with total page 174 pages. Available in PDF, EPUB and Kindle. Book excerpt: Studies were conducted to establish the mechanical properties of uranium 10 wt.% molybdenum (U-10Mo) in both the un-irradiated condition and after neutron irradiation. In the un-irradiated condition, mechanical properties were obtained for various temperatures and after the alloy had been wrought processed by rolling into four different rolling conditions. The irradiated mechanical properties were obtained at various fission densities and then the degradation of the mechanical properties from the un-irradiated to irradiated condition evaluated and a correlation with porosity developed. The mechanical properties obtained of the un-irradiated material differed from that previously published in the literature, which was expected due to the differences in thermomechanical processing conditions between the materials evaluated. The mechanical properties degraded as fission density increased as expected, and correlate to the increase of porosity that develops with increasing fission density.
Author : W. W. Schulz
Publisher :
ISBN 13 :
Total Pages : 42 pages
Book Rating : 4.3/5 (91 download)
Download or read book Reprocessing Uranium-molybdenum Alloy Fuels written by W. W. Schulz and published by . This book was released on 1960 with total page 42 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Dissolution of U-Mo-Alloy Scrap from Fuel Fabrication written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : L. M. Ferris
Publisher :
ISBN 13 :
Total Pages : 44 pages
Book Rating : 4.3/5 (91 download)
Download or read book Aqueous Processes for Dissolution of Uranium-molybdenum Alloy Reactor Fuel Elements written by L. M. Ferris and published by . This book was released on 1961 with total page 44 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : W. W. Schulz
Publisher :
ISBN 13 :
Total Pages : 84 pages
Book Rating : 4.3/5 (91 download)
Download or read book Reprocessing of Low-enrichment Uranium-molybdenum Alloy Fuels written by W. W. Schulz and published by . This book was released on 1959 with total page 84 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :
Publisher :
ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Observation on the Irradiation Behavior of U-Mo Alloy Dispersion Fuel written by and published by . This book was released on 2000 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Solvent Extraction for Uranium Molybdenum Alloy Dissolution Flowsheet written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: H-Canyon Engineering requested the Savannah River National Laboratory (SRNL) to perform two solvent extraction experiments using dissolved Super Kukla (SK) material. The SK material is an uranium (U)-molybdenum (Mo) alloy material of 90% U/10% Mo by weight with 20% 235U enrichment. The first series of solvent extraction tests involved a series of batch distribution coefficient measurements with 7.5 vol % tributylphosphate (TBP)/n-paraffin for extraction from 4-5 M nitric acid (HNO3), using 4 M HNO3-0.02 M ferrous sulfamate (Fe(SO3NH2)2) scrub, 0.01 M HNO3 strip steps with particular emphasis on the distribution of U and Mo in each step. The second set of solvent extraction tests determined whether the 2.5 wt % sodium carbonate (Na2CO3) solvent wash change frequency would need to be modified for the processing of the SK material. The batch distribution coefficient measurements were performed using dissolved SK material diluted to 20 g/L (U + Mo) in 4 M HNO3 and 5 M HNO3. In these experiments, U had a distribution coefficient greater than 2.5 while at least 99% of the nickel (Ni) and greater than 99.9% of the Mo remained in the aqueous phase. After extraction, scrub, and strip steps, the aqueous U product from the strip contains nominally 7.48 [mu]g Mo/g U, significantly less than the maximum allowable limit of 800 [mu]g Mo/g U. Solvent washing experiments were performed to expose a 2.5 wt % Na2CO3 solvent wash solution to the equivalent of 37 solvent wash cycles. The low Mo batch distribution coefficient in this solvent extraction system yields only 0.001-0.005 g/L Mo extracted to the organic. During the solvent washing experiments, the Mo appears to wash from the organic.
Download or read book AQUEOUS PROCESSES FOR DISSOLUTION OF URANIUM-MOLYBDENUM ALLOY REACTOR FUEL ELEMENTS. written by and published by . This book was released on 1961 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Methods for dissolving unirradiated uranium-molybdenum alloy reactcr fuels in nitric acid, nitric acid--ferric nitrate, and nitric acid-- phosphoric acid solutions were studied on a laboratory scale. Flowsheets based on the results propose dissolution of alloys containing 3% molybdenum in boiling 6 M HNO/ sub 3/ to yield stsble solutions that are 0.6 M in uranium and 3 to 4 M in nitric acid. The uranium can then be easily decontaminated and recovered in a conventional Purex-type tributyl phosphate solvent extraction process. Alloys containing 10% molybdenum would be dissolved in boiling 11 M HNO3, allowing molybdic oxide to precipitate. The molybdic oxide, which carries 5-10% of the uranium, is removed by centrifugation and the acidity of the supernatant solution adjusted tc allow recovery of the uranium by Purex-type solvent extraction procedures. The uranium carried by the molybdic oxide is recovered after the MoO/ sub 3/ is dissolved in warm 5 M NaOH. Less than 0.1% of the uranium is solubilized during the caustic dissolution. Alternative methods investigated involve dissolution in nitric acid containing 0.5 to 1 M ferric nitrate to complex the molybdenum. These techniques lead to undesirably large volumes of high-level solvent extraction waste solutions. Phosphate ion is also effective in complexing molybdenum; however, its use in the dissolvent would be purposeless since it must be complexed with iron during solvent extraction. Rates of reaction of the various alloys and the solubility of molybdic oxide were determined in nitric acid, nitric acid-- ferric nitrate, and nitric acid-- phosphonic acid solutions. (auth).
Author :
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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Prototypic Irradiation Testing of High-density U-Mo Alloy Dispersion Fuels written by and published by . This book was released on 2001 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES. written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.
Author :
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ISBN 13 :
Total Pages : 20 pages
Book Rating : 4.:/5 (925 download)
Download or read book Uranium-Molybdenum Dissolution Flowsheet Studies written by and published by . This book was released on 2007 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Super Kukla (SK) Prompt Burst Reactor operated at the Nevada Test Site from 1964 to 1978. The SK material is a uranium-molybdenum (U-Mo) alloy material of 90% U/10% Mo by weight at approximately 20% 235U enrichment. H-Canyon Engineering (HCE) requested that the Savannah River National Lab (SRNL) define a flowsheet for safely and efficiently dissolving the SK material. The objective is to dissolve the material in nitric acid (HNO3) in the H-Canyon dissolvers to a U concentration of 15-20 g/L (3-4 g/L 235U) without the formation of precipitates or the generation of a flammable gas mixture. Testing with SK material validated the applicability of dissolution and solubility data reported in the literature for various U and U-Mo metals. Based on the data, the SK material can be dissolved in boiling 3.0-6.0 M HNO3 to a U concentration of 15-20 g/L and a corresponding Mo concentration of 1.7-2.2 g/L. The optimum flowsheet will use 4.0-5.0 M HNO3 for the starting acid. Any nickel (Ni) cladding associated with the material will dissolve readily. After dissolution is complete, traditional solvent extraction flowsheets can be used to recover and purify the U. Dissolution rates for the SK material are consistent with those reported in the literature and are adequate for H-Canyon processing. When the SK material dissolved at 70-100 o C in 1-6 M HNO3, the reaction bubbled vigorously and released nitrogen oxide (NO) and nitrogen dioxide (NO2) gas. Gas generation tests in 1 M and 2 M HNO3 at 100 o C generated less than 0.1 volume percent hydrogen (H2) gas. It is known that higher HNO3 concentrations are less favorable for H2 production. All tests at 70-100 o C produced sufficient gas to mix the solutions without external agitation. At room temperature in 5 M HNO3, the U-Mo dissolved slowly and the U-laden solution sank to the bottom of the dissolution vessel because of its greater density. The effect of the density difference insures that the SK material cannot dissolve and concentrate within the charge bundles. Solubility behavior of the SK material during dissolution at 70 o C reflected data reported in the literature for 100 o C. When solutions containing solids at 70 o C were heated to 105 o C, the solids dissolved. After 21 days, the samples that had been heated closely resembled the non-heated ones with respect to solids content. Super-saturated solutions of U-Mo have been produced which can be stable for more than 10 days, but these conditions are outside of the bounds of the recommended flowsheet. It is not known how the different dissolution pathways affect solution stability, but the results agree with the fact that solubility should not be affected by the dissolution pathway. Therefore, the literature data should be used as the bounding condition for solubility. Dissolution of the SK material consumed 2.8-8.0 moles of acid per mole of metal dissolved, which agrees with behavior reported elsewhere for U and U-Mo metals. The acid consumption values confirmed that a starting acid concentration in the dissolver of 4.0-5.0 M HNO3 will allow H-Canyon Operations to avoid adjusting the feed from the dissolver prior to solvent extraction while providing maximum operating margin for avoiding precipitate formation.
Download or read book REPROCESSING URANIUM-MOLYBDENUM ALLOY FUELS-DISSOLUTION IN CONCENTRATED NITRIC ACID. written by and published by . This book was released on 1960 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Dissolution of U--3 wt.% Mo alloys in 12 to 14 M HNO3 solutions is discussed. Under these conditions 80 to 95% of the molybdenum content is precipitated as white hydrated molybdic oxide. Molybdic oxide precipitates are bulky; centrifuged volumes range from 6 to 17 vol.% for dissolver solutions 1.0 to 1.5:(M U. The precipitates retain some plutonium and uranium even after thorough washing with water and 1 M HNO3. Plutonium and uranium values in washed precipitates can be recovered by successive treatment of solid residues with caustic and nitric acid. Removal of nitric acid from uranium-molybdenum alloy dissolver solutions by boil-down procedures and by reaction with formaldehyde is discussed. Solutions obtained after removal of nitric acid constitute satisfactory low-acid Redox process feedstocks. (auth).
Author :
Publisher :
ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Irradiation Testing of Intermediate and High-density U-Mo Alloy Dispersion Fuels to High Burnup written by and published by . This book was released on 2000 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book REPROCESSING OF LOW-ENRICHMENT URANIUM-MOLYBDENUM ALLOY FUELS. written by and published by . This book was released on 1959 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Procedures for the dissolution of U-Mo alloy fuels to prepare feed solutions for low-acid (Redox) type solvent extraction processing are presented. U-Mo alloys can be dissolved in boiling ferric nitrate--ritric acid solutions to higher terminal urarium concentrations and lower terminal acidities without precipitation of uranyl molybdate than in nitric acid alone. Anion resin exchange studies indicate the presence of negatively charged iron-molybdenum complex ions in the solutions. The U-Mo alloys also dissolve more rapidly in ferric nitrate--nitnic acid solutions than in nitric acid alone; dissolution rate data are given. Curves delineating free acid, uranium, snd iron (III) concentrations within which solutions stable towards solids formation can be prepared from U-3 wt.% Mo and U-10 wt.% Mo alloys are presented. Stability during prolonged storage of uranium--molybdenum-ferric nitrate--nitric acid solutions is discussed. Data on the oxidation of plutonium in these solutions and on further neutralization of the solutions are presented. Fission product decontamination and product recovery obtained in solvent extraction studies simulating the Redox process are discussed. (auth).
Author : R. M. Willard
Publisher :
ISBN 13 :
Total Pages : 0 pages
Book Rating : 4.:/5 (682 download)
Download or read book Irradiation Swelling, Phase Reversion, and Intergranualr Cracking of U-10 Wt % Mo Fuel Alloy written by R. M. Willard and published by . This book was released on 1965 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :
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ISBN 13 :
Total Pages : 16 pages
Book Rating : 4.:/5 (683 download)
Download or read book Experimental and Calculated Swelling Behavior of U-10 Wt.% Mo Under Low Irradiation Temperatures written by and published by . This book was released on 1998 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: SEM micrographs of U-10 wt.% Mo irradiated at low temperature in the ATR to about 40 at. % burnup show the presence of cavities. We have used a rate-theory-based model to investigate the nucleation and growth of cavities during low-temperature irradiation of uranium-molybdenum alloys in the presence of irradiation-induced interstitial-loop formation and growth. In addition, the evolution of forest dislocations was calculated based on dislocation loop growth and simultaneous climb and glide of unfaded loops. Consolidation of the dislocation structure takes into account capture of interstitial dislocation loops and annihilation of adjacent dislocations, as well as loss to grain boundaries. A di-interstitial is assumed to be the nucleus of a dislocation loop. Cavities are nucleated when two gas atoms come together in the presence of at least one vacancy. Cavity growth occurs by the influx of gas atoms and/or vacancies. In turn, the free interstitial concentration, and thus (due to recombination) the free-vacancy concentration, depends on the dislocation density. Bias-driven growth of cavities can lead to substantial swelling of the alloy (void swelling). However, our calculations indicate that the swelling mechanism in the U-10 wt.% Mo alloy at low irradiation temperatures is fission gas driven. The calculations also indicate that the observed bubbles must be associated with a sub-grain structure. Calculated swelling and bubble-size-distribution are compared with irradiation data.
