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Uranium oxide activation cost study

Description: Owing to continuous uranyl nitrate calcining facilities at Hanford, the reactivity of the A oxide produced was reduced, with lower conversion to UF{sub 4}. A process design was developed for increasing the A{sub 3}O{sub 8} activity by fluidized-bed reduction to UO{sub 2} and reoxidation to A{sub 3}O{sub 8}. Justification for the installation would be in the $250,000 to $300,000 saved in not needing additional fluorine cells and $310,000 per year operating cost savings. The installation would cost $700,000 or $860,000; operating cost would be $270,000 per year. It was concluded that a full-scale process cannot be justified at the time of the study.
Date: April 23, 1957
Creator: McKee, R. W.
Partner: UNT Libraries Government Documents Department

Shuffler bias corrections using calculated count rates

Description: Los Alamos National Laboratory has two identical shufflers that have been calibrated with a dozen U{sub 3}O{sub 8} certified standards from 10 g {sup 235}U to 3600 g {sup 235}U. The shufflers are used to assay a wide variety of material types for their {sup 235}U contents. When the items differ greatly in chemical composition or shape from the U{sub 3}O{sub 8} standards a bias is introduced because the calibration is not appropriate. Recently a new tool has been created to calculate shuffler count rates accurately, and this has been applied to generate bias correction factors. The tool has also been used to verify the masses and count rates of some uncertified U{sub 3}O{sub 8} standards up to 8.0 kg of {sup 235}U which were used to provisionally extend the calibration beyond the 3.6 kg of {sup 235}U mass when a special need arose. Metallic uranium has significantly different neutronic properties from the U{sub 3}O{sub 8} standards and measured count rates from metals are biased low when the U{sub 3}O{sub 8} calibration is applied. The application of the calculational tool to generate bias corrrections for assorted metals will be described. The accuracy of the calculational tool was verified using highly enriched metal disk standards that could be stacked to form cylinders or put into spread arrays.
Date: April 1, 2001
Creator: Rinard, Phillip M.; Hurd, J. R. (Jon R.) & Hsue, F. (Faye)
Partner: UNT Libraries Government Documents Department

Comparison and modeling of aqueous dissolution rates of various uranium oxides

Description: Purpose of this work was to measure and model the intrinsic dissolution rates of U oxides under a variety of well-controlled conditions that are relevant to a geologic repository. When exposed to air at elevated temperature, spent fuel may form the stable phase U{sub 3}O{sub 8}. Dehydrated schoepite, UO{sub 3}{center_dot}H{sub 2}O, exists in drip tests on spent fuel. Equivalent sets of U{sub 3}O{sub 8} and UO{sub 3}{center_dot}H{sub 2}O dissolution experiments allowed a systematic examination of the effects of temperature (25-75 C), pH(8-10), and carbonate (2-200x10{sup -4}molar) concentrations at atmospheric oxygen conditions. Results indicate that UO{sub 3}{center_dot}H{sub 2}O has a much higher dissolution rate (at least tenfold) than U{sub 3}O{sub 8} under the same conditions. The intrinsic dissolution rate of unirradiated U{sub 3}O{sub 8} is about twice that of UO{sub 2}. Dissolution of both U{sub 3}O{sub 8} and UO{sub 3}{center_dot}H{sub 2}O shows a very high sensitivity to carbonate concentration. Present results show a 25 to 50-fold increase in room-temperature UO{sub 3}{center_dot}H{sub 2}O dissolution rates between the highest and lowest carbonate concentrations. As with the UO{sub 2} dissolution data, the classical observed chemical kinetic rate law was used to model the U{sub 3}O{sub 8} dissolution rate data. The pH did not have much effect on the models, in agreement with earlier analysis of the UO{sub 2} and spent fuel dissolution data. However, carbonate concentration, not temperature, had the strongest effect on the U{sub 3}O{sub 8} dissolution rate. The U{sub 3}O{sub 8} dissolution activation energy was about 6000 cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and UO{sub 2}, respectively.
Date: November 1, 1996
Creator: Steward, S.A.
Partner: UNT Libraries Government Documents Department

Head-end reprocessing studies of H.B. Robinson-2 fuel: II. Parametric voloxidation studies

Description: A series of hot-cell tests was conducted with UO{sub 2} that had been irradiated to an average of 28,000 MWd/t in the H.B. Robinson-2 reactor of the Carolina Power and Light Company. The tests examined the effects of temperature and of the rate of oxygen supply on the release of gaseous and semivolatile fission products, while the fuel fragments were tumbled at 12 rpm during voloxidation - the high-temperature oxidation of UO{sub 2} to U{sub 3}O{sub 8}. The experiments showed that >99.9% of the tritium in the irradiated UO{sub 2} was released to the off-gas stream at temperatures of 480 and 550{sup 0}C and at oxygen feed rates ranging from 0.1 to 1.2 mol/h. The release of {sup 85}Kr varied from 2 to 7% of the fuel inventory. The U{sub 3}O{sub 8} product ({similar_to}99% smaller than 44 {mu}m) was easily dissolved in 7 M HNO{sub 3}. One 2-h leach in 7 M HNO{sub 3} dissolved {similar_to}99.5% of the heavy metals; a second 2-h leach in 7 M HNO{sub 3} brought the total to >99.98%. Voloxidation did not affect the final solubility of the uranium and plutonium but did increase the weight of the insoluble fission product residue from 0.18% of the irradiated UO{sub 2} to {similar_to}0.62%.
Date: May 1, 1980
Creator: Goode, J.H.; Stacy, R.G. & Vaughen, V.C.A.
Partner: UNT Libraries Government Documents Department

Use of Cation Exchange Resins for Production of U{sub 3}O{sub 8} Suitable for the Al-U{sub 3}O{sub 8} Powder Metallurgy Process

Description: This report describes the production of U{sub 3}O{sub 8} powders from three types of cation exchange resins: Dowex 50W, a strong acid, sulfonate resin; AG MP-50, a macroporous form of sulfonate resin; and Bio-Rex 70, a weak acid, carboxylic resin.
Date: September 17, 2001
Creator: Mosley, W.C.
Partner: UNT Libraries Government Documents Department

Uranium purchases report 1994

Description: US utilities are required to report to the Secretary of Energy annually the country of origin and the seller of any uranium or enriched uranium purchased or imported into the US, as well as the country of origin and seller of any enrichment services purchased by the utility. This report compiles these data and also contains a glossary of terms and additional purchase information covering average price and contract duration. 3 tabs.
Date: July 1, 1995
Partner: UNT Libraries Government Documents Department

CNEA/ANL collaboration program to develop an optimized version of DART validation and assessment by means of U{sub 3}Si{sub x} and U{sub 3}O{sub 8-}Al dispersed CNEA miniplate irradiation behavior.

Description: The DART code is based upon a thermomechanical model that can predict swelling, recrystallization, fuel-meat interdiffusion and other issues related with MTR dispersed FE behavior under irradiation. As a part of a common effort to develop an optimized version of DART, a comparison between DART predictions and CNEA miniplates irradiation experimental data was made. The irradiation took place during 1981-82 for U3O8 miniplates and 1985-86 for U{sub 3}Si{sub x} at Oak Ridge Research Reactor (ORR). The microphotographs were studied by means of IMAWIN 3.0 Image Analysis Code and different fission gas bubbles distributions were obtained. Also it was possible to find and identify different morphologic zones. In both kinds of fuels, different phases were recognized, like particle peripheral zones with evidence of Al-U reaction, internal recrystallized zones and bubbles. A very good agreement between code prediction and irradiation results was found. The few discrepancies are due to local, fabrication and irradiation uncertainties, as the presence of U{sub 3}Si phase in U{sub 3}Si{sub 2} particles and effective burnup.
Date: October 16, 1998
Creator: Solis, D.
Partner: UNT Libraries Government Documents Department

Redox Bias in Loss on Ignition Moisture Measurement for Relatively Pure Plutonium-Bearing Oxide Materials

Description: This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD- 3013). An immediate application is to Rocky Flats (RF) materials derived from high-grade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidation/reduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation show s that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LOI stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Significant bias also requires that UO2 components remain largely unoxidized after calcination and are largely converted to U3O8 during LOI testing at only slightly higher temperatures. Based on well-established literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LOI weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confirm these expectations and to provide a more authoritative basis for bounding LOI oxidation/reduction biases. LOI bias evaluation is more difficult for lower purity materials and for fuel-type uranium-plutonium oxides. However, even in these cases ...
Date: February 26, 2002
Creator: Eller, P. G.; Stakebake, J. L. & Cooper, T. D.
Partner: UNT Libraries Government Documents Department

Design Parameters for a Natural Uranium UO{sub 3} or U{sub 3}O{sub 8} Fueled Nuclear Reactor

Description: A recent Oak Ridge National Laboratory report provided preliminary analyses to propose alternative design parameters for a nuclear reactor that could be fueled with natural UO{sub 3} or U{sub 3}O{sub 8} and moderated with either heavy water or reactor-grade graphite. This report provides more specific reactor design and operating parameters for a heavy water-moderated reactor only. The basic assumptions and analytical approach are discussed together with the results of the analysis.
Date: November 15, 2002
Creator: Hopper, C.M.
Partner: UNT Libraries Government Documents Department

Dissolving uranium oxide--aluminum fuel

Description: The dissolution of aluminum-clad uranium oxide-aluminum fuel was studied to provide basic data for dissolving this type of enriched uranium fuel at the Savannah River Plant. The studies also included the dissolution of a similar material prepared from scrap uranium oxides that were to be recycled through the solvent extraction process. The dissolving behavior of uranium oxide-aluminum core material is similar to that of U-Al alloy. Dissolving rates are rapid in HNO/sub 3/-Hg(NO/sub 3/)/sub 2/ solutions. Irradiation reduce s the dissolving rate and increases mechanical strength. A dissolution model for use in nuclear safety analyses is developed, . based on the observed dissolving characteristics. (auth)
Date: November 1, 1973
Creator: Perkins, W.C.
Partner: UNT Libraries Government Documents Department

Effect of Process Variables During the Head-End Treatment of Spent Oxide Fuel

Description: The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. The effects of temperature, pressure, oxidative gas, and cladding have been studied with irradiated spent oxide fuel to determine the optimum conditions for process control. Experiments with temperatures ranging from 500oC to 1250oC have been performed on spent fuel using either air or oxygen gas for the oxidative reaction. Various flowrates and applications have been tested with the oxidative gases to discern the effects on the process. Tests have also been performed under vacuum conditions, following the oxidation cycle, at high temperatures to improve the removal of fission products. The effects of cladding on fission product removal have also been investigated with released fuel under vacuum and high temperature conditions. Results from these experiments will be presented as well as operating conditions based on particle size and decladding characteristics.
Date: August 1, 2006
Creator: Bateman, K.J.; Morgan, C.D.; Berg, J.F.; Brough, D.J.; Crane, P.J.; Cummings, D.G. et al.
Partner: UNT Libraries Government Documents Department

Electrolytic Reduction of Spent Nuclear Oxide Fuel -- Effects of Fuel Form and Cathode Containment Materials on Bench-Scale Operations

Description: A collaborative effort between the Idaho National Laboratory (INL) and Korea Atomic Energy Research Institute (KAERI) is underway per an International Nuclear Energy Research Initiative to advance the development of a pyrochemical process for the treatment of spent nuclear oxide fuel. To assess the effects of specific process parameters that differ between oxide reduction operations at INL and KAERI, a series of 4 electrolytic reduction runs will be performed with a single salt loading of LiCl-Li2O at 650 °C using a test apparatus located inside of a hot cell at INL. The spent oxide fuel for the tests will be irradiated UO2 that has been subjected to a voloxidation process to form U3O8. The primary variables in the 4 electrolytic reduction runs will be fuel basket containment material and Li2O concentration in the LiCl salt. All 4 runs will be performed with comparable fuel loadings (approximately 50 g) and fuel compositions and will utilize a platinum anode and a Ni/NiO reference electrode. The first 2 runs will elucidate the effect of fuel form on the electrolytic reduction process by comparison of the above test results with U3O8 versus results from previous tests with UO2. The first 3 runs will investigate the impact that the cathode containment material has on the electrolytic reduction of spent oxide fuel. The 3rd and 4th runs will investigate the effect of Li2O concentration on the reduction process with a porous MgO cathode containment.
Date: September 1, 2007
Creator: Herrmann, S. D.
Partner: UNT Libraries Government Documents Department

Fission Product Removal From Spent Oxide Fuel By Head-End Processing

Description: The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. Experiments have been performed with irradiated oxide fuel to evaluate the removal of fission products. During these experiments, operating parameters such as temperature and pressure have been varied to discern their effects on the behavior of specific fission products. In general, the extent of removal increases with increasing operating temperature and decreasing pressure. Removal efficiencies as high as 98% have been achieved during testing. Given the results of testing, an explanation of the likely fission product species being removed during the test program is also provided. In addition, experiments have been performed with other oxidative gases (steam and ozone) on surrogates to determine their potential benefit for removal of fission products.
Date: October 1, 2005
Creator: Westphal, B. R.; Bateman, K. J.; Lind, R. P.; Howden, K. L. & Cul, G. D. Del
Partner: UNT Libraries Government Documents Department

Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

Description: A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.
Date: July 1, 2008
Creator: Westphal, B.R.; Park, J.J.; Shin, J.M.; Park, G.I.; Bateman, K.J. & Wahlquist, D.L.
Partner: UNT Libraries Government Documents Department

Comparison of the NDA of HEU Oxide between the AWCC and the HPGe Detector

Description: This paper compares the performance of the Active Well Coincidence Counter (AWCC) with the performance of high resolution gamma spectrometry using an HPGe detector to nondestructively assay highly enriched (HEU) oxide. Traditionally the AWCC was considered to be the more appropriate instrument for this measurement. Although the AWCC had a high degree of precision, the HPGe provided the more accurate measurement of this material. The AWCC determines mass of U-235 from the coincident pairs of neutron detections, or doubles rate. The HPGe determines the mass of both U-235 and U238, the enrichment, and the quantity of other radioisotopes. The Tl-208 gamma rays were used to verify the amount of attenuation for the HPGe analysis. Fifty-four cans of enriched U3O8 were shipped to the Y-12 National Security Complex from Los Alamos National Laboratory (LANL) under Scrap Declaration LANL-45. The declared values for net weight, mass of uranium, mass of U-235, and enrichment (percent mass of U-235 to total uranium) are shown in Table A-1. The masses of U-235 range from 104g to 2404g and the enrichment varies from 20% to 98%.
Date: December 1, 2009
Creator: Chiang, L. G.; Oberer, R. B.; Gunn, C. A.; Dukes, E. E. & Akin, J. A.
Partner: UNT Libraries Government Documents Department

A Technique to Determine Billet Core Charge Weight for P/M Fuel Tubes

Description: The core length in an extruded tube depends on the weight of powder in the billet core. In the past, the amount of aluminum powder needed to give a specified core length was determined empirically. This report gives a technique for calculating the weight of aluminum powder for the P/M core. An equation has been derived which can be used to determine the amount of aluminum needed for P/M billet core charge weights. Good agreement was obtained when compared to Mark 22 tube extrusion data. From the calculated charge weight, the elastomeric bag can be designed and made to compact the U3O8-Al core.
Date: July 2, 2001
Creator: Peacock, H.B.
Partner: UNT Libraries Government Documents Department

Bounding criticality safety analyses for shipments of unconfigured spent nuclear fuel

Description: In November 1996, a request was made to the US Department of Energy for a waiver for three shipments of spent nuclear fuel (SNF) from Oak Ridge National Laboratory (ORNL) to the Savannah River Site (SRS) in the US NRC certified BMI-1 cask (CoC 5957). Although the post-irradiation fissile mass (based on chemical assays) in each shipment was less than 800 g, a criticality safety analysis was needed because the pre-irradiation mass exceeded 800 g, the fissile material limit in the CoC. The analyses were performed on SNF consisting of aluminum-clad U{sub 3}O{sub 8}, UAl{sub x}, and U{sub 3}Si{sub 2} plates, fragments and pieces that had been irradiated at ORNL during the Reduced Enrichment Research and Test Reactor Program of the 1980s. The highlights of the approach used to analyze this unique SNF and the benefits of the waiver are presented in this paper.
Date: June 1, 1998
Creator: Lichtenwalter, J.J. & Parks, C.V.
Partner: UNT Libraries Government Documents Department