Production of Low Enriched Uranium Nitride Kernels for TRISO Particle Irradiation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMurray, J. W.; Silva, C. M.; Helmreich, G. W.

2016-06-01

A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC 0.27N 0.73. The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% formore » a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.« less

Data Compilation for AGR-3/4 Designed-to-Fail (DTF) Fuel Particle Batch LEU04-02DTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D; Miller, James Henry

2008-10-01

This document is a compilation of coating and characterization data for the AGR-3/4 designed-to-fail (DTF) particles. The DTF coating is a high density, high anisotropy pyrocarbon coating of nominal 20 {micro}m thickness that is deposited directly on the kernel. The purpose of this coating is to fail early in the irradiation, resulting in a controlled release of fission products which can be analyzed to provide data on fission product transport. A small number of DTF particles will be included with standard TRISO driver fuel particles in the AGR-3 and AGR-4 compacts. The ORNL Coated Particle Fuel Development Laboratory 50-mm diametermore » fluidized bed coater was used to coat the DTF particles. The coatings were produced using procedures and process parameters that were developed in an earlier phase of the project as documented in 'Summary Report on the Development of Procedures for the Fabrication of AGR-3/4 Design-to-Fail Particles', ORNL/TM-2008/161. Two coating runs were conducted using the approved coating parameters. NUCO425-06DTF was a final process qualification batch using natural enrichment uranium carbide/uranium oxide (UCO) kernels. After the qualification run, LEU04-02DTF was produced using low enriched UCO kernels. Both runs were inspected and determined to meet the specifications for DTF particles in section 5 of the AGR-3 & 4 Fuel Product Specification (EDF-6638, Rev.1). Table 1 provides a summary of key properties of the DTF layer. For comparison purposes, an archive sample of DTF particles produced by General Atomics was characterized using identical methods. This data is also summarized in Table 1.« less

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Na, Chongzheng

2016-10-17

Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety ofmore » synthetic methods, we show that synthesizing graphene-­supported UO 2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO 2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO 2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO 2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.« less

Growth of the interaction layer around fuel particles in dispersion fuel

NASA Astrophysics Data System (ADS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.

Characterisation and dissolution of depleted uranium aerosols produced during impacts of kinetic energy penetrators against a tank.

PubMed

Chazel, V; Gerasimo, P; Dabouis, V; Laroche, P; Paquet, F

2003-01-01

Aerosols produced during impacts of depleted uranium (DU) penetrators against the glacis (sloping armour) and the turret of a tank were sampled. The concentration and size distribution were determined. Activity median aerodynamic diameters were 1 microm (geometric standard deviation, sigma(g) = 3.7) and 2 microm (sigma(g) = 2.5), respectively, for glacis and turret. The mean air concentration was 120 Bq m(-3), i.e. 8.5 mg m(-3) of DU. Filters analysed by scanning electron microscopy (SEM) and X ray diffraction showed two types of particles (fine particles and large molten particles) composed mainly of a mixture of uranium and aluminium. The uranium oxides were mostly U3O8, UO2.25 and probably UO3.01 and a mixed compound of U and Al. The kinetics of dissolution in three media (HCO3-, HCl and Gamble's solution) were determined using in-vitro tests. The slow dissolution rates were respectively slow, and intermediate between slow and moderate, and the rapid dissolution fractions were mostly intermediate between moderate and fast. According to the in-vitro results for Gamble's solution, and based on a hypothetical single acute inhalation of 90 Bq, effective doses integrated up to 1 y after incorporation were 0.54 and 0.56 mSv, respectively, for aerosols from glacis and turret. In comparison, the ICRP limits are 20 mSv y(-1) for workers and 1 mSv y(-1) for members of the public. A kidney concentration of approximately 0.1 microg U g(-1) was predicted and should not, in this case, lead to kidney damage.

Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report March - October 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kips, R; Kristo, M; Hutcheon, I

2009-11-22

Nuclear forensics relies on the analysis of certain sample characteristics to determine the origin and history of a nuclear material. In the specific case of uranium enrichment facilities, it is the release of trace amounts of uranium hexafluoride (UF{sub 6}) gas - used for the enrichment of uranium - that leaves a process-characteristic fingerprint. When UF{sub 6} gas interacts with atmospheric moisture, uranium oxyfluoride particles or particle agglomerates are formed with sizes ranging from several microns down to a few tens of nanometers. These particles are routinely collected by safeguards organizations, such as the International Atomic Energy Agency (IAEA), allowingmore » them to verify whether a facility is compliant with its declarations. Spectrometric analysis of uranium particles from UF{sub 6} hydrolysis has revealed the presence of both particles that contain fluorine, and particles that do not. It is therefore assumed that uranium oxyfluoride is unstable, and decomposes to form uranium oxide. Understanding the rate of fluorine loss in uranium oxyfluoride particles, and the parameters that control it, may therefore contribute to placing boundaries on the particle's exposure time in the environment. Expressly for the purpose of this study, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (EU-JRC-IRMM) from a static release of UF{sub 6} in a humid atmosphere. The majority of the samples was stored in controlled temperature, humidity and lighting conditions. Single particles were characterized by a suite of micro-analytical techniques, including NanoSIMS, micro-Raman spectrometry (MRS), scanning (SEM) and transmission (TEM) electron microscopy, energy-dispersive X-ray spectrometry (EDX) and focused ion beam (FIB). The small particle size was found to be the main analytical challenge. The relative amount of fluorine, as well as the particle chemical composition and morphology were determined at different stages in the ageing process, and immediately after preparation. This report summarizes our most recent findings for each of the analytical techniques listed above, and provides an outlook on what remains to be resolved. Additional spectroscopic and mass spectrometric measurements were carried out at Pacific Northwest National Laboratory, but are not included in this summary.« less

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Image segmentation for uranium isotopic analysis by SIMS: Combined adaptive thresholding and marker controlled watershed approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Willingham, David G.; Naes, Benjamin E.; Heasler, Patrick G.

A novel approach to particle identification and particle isotope ratio determination has been developed for nuclear safeguard applications. This particle search approach combines an adaptive thresholding algorithm and marker-controlled watershed segmentation (MCWS) transform, which improves the secondary ion mass spectrometry (SIMS) isotopic analysis of uranium containing particle populations for nuclear safeguards applications. The Niblack assisted MCWS approach (a.k.a. SEEKER) developed for this work has improved the identification of isotopically unique uranium particles under conditions that have historically presented significant challenges for SIMS image data processing techniques. Particles obtained from five NIST uranium certified reference materials (CRM U129A, U015, U150, U500more » and U850) were successfully identified in regions of SIMS image data 1) where a high variability in image intensity existed, 2) where particles were touching or were in close proximity to one another and/or 3) where the magnitude of ion signal for a given region was count limited. Analysis of the isotopic distributions of uranium containing particles identified by SEEKER showed four distinct, accurately identified 235U enrichment distributions, corresponding to the NIST certified 235U/238U isotope ratios for CRM U129A/U015 (not statistically differentiated), U150, U500 and U850. Additionally, comparison of the minor uranium isotope (234U, 235U and 236U) atom percent values verified that, even in the absence of high precision isotope ratio measurements, SEEKER could be used to segment isotopically unique uranium particles from SIMS image data. Although demonstrated specifically for SIMS analysis of uranium containing particles for nuclear safeguards, SEEKER has application in addressing a broad set of image processing challenges.« less

Design of a uranium-dioxide powder spheroidization system by plasma processing

NASA Astrophysics Data System (ADS)

Cavender, Daniel

The plasma spheroidization system (PSS) is the first process in the development of a tungsten-uranium dioxide (W-UO2) ceramic-metallic (cermet) fuel for nuclear thermal rocket (NTR) propulsion. For the purposes of fissile fuel retention, UO2 spheroids ranging in size from 50 - 100 micrometers (μm) in diameter will be encapsulated in a tungsten shell. The PSS produces spherical particles by melting angular stock particles in an argon-hydrogen plasma jet where they become spherical due to surface tension. Surrogate CeO 2 powder was used in place of UO2 for system and process parameter development. Stock and spheroidized powders were micrographed using optical and scanning electron microscopy and evaluated by statistical methods to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders showed a statistically significant improvement in spherocity, with greater that 60% of the examined particles having an irregularity parameter of equal to or lower than 1.2, compared to stock powder.

Processing of uranium oxide and silicon carbide based fuel using polymer infiltration and pyrolysis

NASA Astrophysics Data System (ADS)

Singh, Abhishek K.; Zunjarrao, Suraj C.; Singh, Raman P.

2008-09-01

Ceramic composite pellets consisting of uranium oxide, UO 2, contained within a silicon carbide matrix, were fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, particles of depleted uranium oxide, in the form of U 3O 8, were dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900 °C under a continuous flow of ultra high purity argon. The pyrolysis of AHPCS, at these temperatures, produced near-stoichiometric amorphous silicon carbide ( a-SiC). Multiple polymer infiltration and pyrolysis (PIP) cycles were performed to minimize open porosity and densify the silicon carbide matrix. Analytical characterization was conducted to investigate chemical interaction between U 3O 8 and SiC. It was observed that U 3O 8 reacted with AHPCS during the very first pyrolysis cycle, and was converted to UO 2. As a result, final composition of the material consisted of UO 2 particles contained in an a-SiC matrix. The physical and mechanical properties were also quantified. It is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix.

Numerical simulation of migration behavior of uranium ore dust particles in the human respiratory tract

NASA Astrophysics Data System (ADS)

Ye, Yong-jun; Yin, An-song; Li, Zhi; Lei, Bo; Ding, De-xin

2017-04-01

There is a certain concentration of radioactive dust particles in the air of workplace of underground uranium mines. Some small diameter particles will pass through the masks and enter the respiratory tract which will cause radiation damage to the human body. In order to study deposition regularity of uranium dust in the human respiratory tract, in this paper, we firstly use the RNG turbulence model to simulate the gas flow field in the human respiratory tract Z0 ∼ Z3 level under different respiratory intensity. Then we use DPM discrete phase model to simulate the concentration, particle size distribution, deposition rate and deposition share of uranium dust particles after being filtered through the masks in the human respiratory tract Z0 to Z3 bronchus. According to the simulation results, we have got the following conclusions: the particles’ number concentration of uranium dust after being filtered through the mask in the human respiratory tract basically decreases with the increasing of particle size under different respiratory intensities on the environment of uranium mine. In addition, the intensity of respiration and the mass concentration of particles have an important influence on the deposition rate and the deposition of particles in the respiratory tract.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jolly, Brian C.; Helmreich, Grant; Cooley, Kevin M.

In support of fully ceramic microencapsulated (FCM) fuel development, coating development work is ongoing at Oak Ridge National Laboratory (ORNL) to produce tri-structural isotropic (TRISO) coated fuel particles with both UN kernels and surrogate (uranium-free) kernels. The nitride kernels are used to increase fissile density in these SiC-matrix fuel pellets with details described elsewhere. The surrogate TRISO particles are necessary for separate effects testing and for utilization in the consolidation process development. This report focuses on the fabrication and characterization of surrogate TRISO particles which use 800μm in diameter ZrO 2 microspheres as the kernel.

Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

NASA Astrophysics Data System (ADS)

Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

2014-05-01

The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

Simultaneous determination of the quantity and isotopic ratios of uranium in individual micro-particles by isotope dilution thermal ionization mass spectrometry (ID-TIMS).

PubMed

Park, Jong-Ho; Choi, Eun-Ju

2016-11-01

A method to determine the quantity and isotopic ratios of uranium in individual micro-particles simultaneously by isotope dilution thermal ionization mass spectrometry (ID-TIMS) has been developed. This method consists of sequential sample and spike loading, ID-TIMS for isotopic measurement, and application of a series of mathematical procedures to remove the contribution of uranium in the spike. The homogeneity of evaporation and ionization of uranium content was confirmed by the consistent ratio of n((233)U)/n((238)U) determined by TIMS measurements. Verification of the method was performed using U030 solution droplets and U030 particles. Good agreements of resulting uranium quantity, n((235)U)/n((238)U), and n((236)U)/n((238)U) with the estimated or certified values showed the validity of this newly developed method for particle analysis when simultaneous determination of the quantity and isotopic ratios of uranium is required. Copyright © 2016 Elsevier B.V. All rights reserved.

Preparation of UC0.07-0.10N0.90-0.93 spheres for TRISO coated fuel particles

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, C. M.; Lindemer, T. B.; Johnson, J. A.; Collins, J. L.

2014-05-01

The US Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with dense uranium nitride (UN) kernels with diameters of 650 or 800 μm. The objectives of this effort are to make uranium oxide microspheres with adequately dispersed carbon nanoparticles and to convert these microspheres into UN spheres, which could be then sintered into kernels. Recent improvements to the internal gelation process were successfully applied to the production of uranium gel spheres with different concentrations of carbon black. After the spheres were washed and dried, a simple two-step heat profile was used to produce porous microspheres with a chemical composition of UC0.07-0.10N0.90-0.93. The first step involved heating the microspheres to 2023 K in a vacuum, and in the second step, the microspheres were held at 1873 K for 6 h in flowing nitrogen.

Nuclear Rocket Ceramic Metal Fuel Fabrication Using Tungsten Powder Coating and Spark Plasma Sintering

NASA Technical Reports Server (NTRS)

Barnes, M. W.; Tucker, D. S.; Hone, L.; Cook, S.

2017-01-01

Nuclear thermal propulsion is an enabling technology for crewed Mars missions. An investigation was conducted to evaluate spark plasma sintering (SPS) as a method to produce tungsten-depleted uranium dioxide (W-dUO2) fuel material when employing fuel particles that were tungsten powder coated. Ceramic metal fuel wafers were produced from a blend of W-60vol% dUO2 powder that was sintered via SPS. The maximum sintering temperatures were varied from 1,600 to 1,850 C while applying a 50-MPa axial load. Wafers exhibited high density (>95% of theoretical) and a uniform microstructure (fuel particles uniformly dispersed throughout tungsten matrix).

Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

2016-08-01

Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO 2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where somemore » of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.« less

Rapid removal of uranium from aqueous solutions using magnetic Fe3O4@SiO2 composite particles.

PubMed

Fan, Fang-Li; Qin, Zhi; Bai, Jing; Rong, Wei-Dong; Fan, Fu-You; Tian, Wei; Wu, Xiao-Lei; Wang, Yang; Zhao, Liang

2012-04-01

Rapid removal of U(VI) from aqueous solutions was investigated using magnetic Fe(3)O(4)@SiO(2) composite particles as the novel adsorbent. Batch experiments were conducted to study the effects of initial pH, amount of adsorbent, shaking time and initial U(VI) concentrations on uranium sorption efficiency as well as the desorbing of U(VI). The sorption of uranium on Fe(3)O(4)@SiO(2) composite particles was pH-dependent, and the optimal pH was 6.0. In kinetics studies, the sorption equilibrium can be reached within 180 min, and the experimental data were well fitted by the pseudo-second-order model, and the equilibrium sorption capacities calculated by the model were almost the same as those determined by experiments. The Langmuir sorption isotherm model correlates well with the uranium sorption equilibrium data for the concentration range of 20-200 mg/L. The maximum uranium sorption capacity onto magnetic Fe(3)O(4)@SiO(2) composite particles was estimated to be about 52 mg/g at 25 °C. The highest values of uranium desorption (98%) was achieved using 0.01 M HCl as the desorbing agent. Fe(3)O(4)@SiO(2) composite particles showed a good selectivity for uranium from aqueous solution with other interfering cation ions. Present study suggested that magnetic Fe(3)O(4)@SiO(2) composite particles can be used as a potential adsorbent for sorption uranium and also provided a simple, fast separation method for removal of heavy metal ion from aqueous solution. Copyright © 2011 Elsevier Ltd. All rights reserved.

Method of preparation of uranium nitride

DOEpatents

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

An aerosol particle containing enriched uranium encountered in the remote upper troposphere.

PubMed

Murphy, D M; Froyd, K D; Apel, E; Blake, D; Blake, N; Evangeliou, N; Hornbrook, R S; Peischl, J; Ray, E; Ryerson, T B; Thompson, C; Stohl, A

2018-04-01

We describe a submicron aerosol particle sampled at an altitude of 7 km near the Aleutian Islands that contained a small percentage of enriched uranium oxide. 235 U was 3.1 ± 0.5% of 238 U. During twenty years of aircraft sampling of millions of particles in the global atmosphere, we have rarely encountered a particle with a similarly high content of 238 U and never a particle with enriched 235 U. The bulk of the particle consisted of material consistent with combustion of heavy fuel oil. Analysis of wind trajectories and particle dispersion model results show that the particle could have originated from a variety of areas across Asia. The source of such a particle is unclear, and the particle is described here in case it indicates a novel source where enriched uranium was dispersed. Published by Elsevier Ltd.

An aerosol particle containing enriched uranium encountered during routine sampling

NASA Astrophysics Data System (ADS)

Murphy, Daniel; Froyd, Karl; Evangeliou, NIkolaos; Stohl, Andreas

2017-04-01

The composition of single aerosol particles has been measured using a laser ionization mass spectrometer during the global Atmospheric Tomography mission. The measurements were targeting the background atmosphere, not radiochemical emissions. One sub-micron particle sampled at about 7 km altitude near the Aleutian Islands contained uranium with approximately 3% 235U. It is the only particle with enriched uranium out of millions of particles sampled over several decades of measurements with this instrument. The particle also contained vanadium, alkali metals, and organic material similar to that present in emissions from combustion of heavy oil. No zirconium or other metals that might be characteristic of nuclear reactors were present, probably suggesting a source other than Fukushima or Chernobyl. Back trajectories suggest several areas in Asia that might be sources for the particle.

ELECTRONUCLEAR REACTOR

DOEpatents

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Physicochemical Characterization of Capstone Depleted Uranium Aerosols II: Particle Size Distributions as a Function of Time

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Yung-Sung; Kenoyer, Judson L.; Guilmette, Raymond A.

2009-03-01

The Capstone Depleted Uranium (DU) Aerosol Study, which generated and characterized aerosols containing depleted uranium from perforation of armored vehicles with large-caliber DU penetrators, incorporated a sampling protocol to evaluated particle size distributions. Aerosol particle size distribution is an important parameter that influences aerosol transport and deposition processes as well as the dosimetry of the inhaled particles. These aerosols were collected on cascade impactor substrates using a pre-established time sequence following the firing event to analyze the uranium concentration and particle size of the aerosols as a function of time. The impactor substrates were analyzed using beta spectrometry, and themore » derived uranium content of each served as input to the evaluation of particle size distributions. Activity median aerodynamic diameters (AMADs) of the particle size distributions were evaluated using unimodal and bimodal models. The particle size data from the impactor measurements was quite variable. Most size distributions measured in the test based on activity had bimodal size distributions with a small particle size mode in the range of between 0.2 and 1.2 um and a large size mode between 2 and 15 um. In general, the evolution of particle size over time showed an overall decrease of average particle size from AMADs of 5 to 10 um shortly after perforation to around 1 um at the end of the 2-hr sampling period. The AMADs generally decreased over time because of settling. Additionally, the median diameter of the larger size mode decreased with time. These results were used to estimate the dosimetry of inhaled DU particles.« less

Selective, high-energy beta scintillation sensor for real-time, in situ characterization of uranium-238 and strontium-90

NASA Astrophysics Data System (ADS)

Schilk, A. J.; Abel, K. H.; Brown, D. P.; Thompson, R. C.; Knopf, M. A.; Hubbard, C. W.

1994-04-01

A novel scintillating-fiber sensor for detecting high-energy beta particles has been designed and built at the Pacific Northwest Laboratory to characterize U-238 and Sr-90 in surface soils. High-energy betas generate unique signals as they pass through multiple layers of scintillating fibers that make up the active region of the detector. Lower-energy beta particles, gamma rays, and cosmic-ray-generated particles comprise the majority of the background interferences. The resulting signals produced by these latter phenomena are effectively discriminated against due to the combination of the sensor's multilayer configuration and its interlayer coincidence/anticoincidence circuitry.

Checking the possibility of controlling fuel element by X-ray computerized tomography

NASA Astrophysics Data System (ADS)

Trinh, V. B.; Zhong, Y.; Osipov, S. P.; Batranin, A. V.

2017-08-01

The article considers the possibility of checking fuel elements by X-ray computerized tomography. The checking tasks are based on the detection of particles of active material, evaluation of the heterogeneity of the distribution of uranium salts and the detection of clusters of uranium particles. First of all, scheme of scanning improve the performance and quality of the resulting three-dimensional images of the internal structure is determined. Further, the possibility of detecting clusters of uranium particles having the size of 1 mm3 and measuring the coordinates of clusters of uranium particles in the middle layer with the accuracy of within a voxel size (for the considered experiments of about 80 μm) is experimentally proved in the main part. The problem of estimating the heterogeneity of the distribution of the active material in the middle layer and the detection of particles of active material with a nominal diameter of 0.1 mm in the “blank” is solved.

40 CFR 141.26 - Monitoring frequency and compliance requirements for radionuclides in community water systems.

Code of Federal Regulations, 2012 CFR

2012-07-01

.... For the purposes of monitoring for gross alpha particle activity, radium-226, radium-228, uranium, and... monitoring: Systems must conduct initial monitoring for gross alpha particle activity, radium-226, radium-228...) For gross alpha particle activity, uranium, radium-226, and radium-228 monitoring, the State may waive...

Removal of toxic uranium from synthetic nuclear power reactor effluents using uranyl ion imprinted polymer particles.

PubMed

Preetha, Chandrika Ravindran; Gladis, Joseph Mary; Rao, Talasila Prasada; Venkateswaran, Gopala

2006-05-01

Major quantities of uranium find use as nuclear fuel in nuclear power reactors. In view of the extreme toxicity of uranium and consequent stringent limits fixed by WHO and various national governments, it is essential to remove uranium from nuclear power reactor effluents before discharge into environment. Ion imprinted polymer (IIP) materials have traditionally been used for the recovery of uranium from dilute aqueous solutions prior to detection or from seawater. We now describe the use of IIP materials for selective removal of uranium from a typical synthetic nuclear power reactor effluent. The IIP materials were prepared for uranyl ion (imprint ion) by forming binary salicylaldoxime (SALO) or 4-vinylpyridine (VP) or ternary SALO-VP complexes in 2-methoxyethanol (porogen) and copolymerizing in the presence of styrene (monomer), divinylbenzene (cross-linking monomer), and 2,2'-azobisisobutyronitrile (initiator). The resulting materials were then ground and sieved to obtain unleached polymer particles. Leached IIP particles were obtained by leaching the imprint ions with 6.0 M HCl. Control polymer particles were also prepared analogously without the imprint ion. The IIP particles obtained with ternary complex alone gave quantitative removal of uranyl ion in the pH range 3.5-5.0 with as low as 0.08 g. The retention capacity of uranyl IIP particles was found to be 98.50 mg/g of polymer. The present study successfully demonstrates the feasibility of removing uranyl ions selectively in the range 5 microg - 300 mg present in 500 mL of synthetic nuclear power reactor effluent containing a host of other inorganic species.

Method for magnesium sulfate recovery

DOEpatents

Gay, Richard L.; Grantham, LeRoy F.

1987-01-01

A method of obtaining magnesium sulfate substantially free from radioactive uranium from a slag containing the same and having a radioactivity level of at least about 7000 pCi/gm. The slag is ground to a particle size of about 200 microns or less. The ground slag is then contacted with a concentrated sulfuric acid under certain prescribed conditions to produce a liquid product and a solid product. The particulate solid product and a minor amount of the liquid is then treated to produce a solid residue consisting essentially of magnesium sulfate substantially free of uranium and having a residual radioactivity level of less than 1000 pCi/gm. In accordance with the preferred embodiment of the invention, a catalyst and an oxidizing agent are used during the initial acid treatment and a final solid residue has a radioactivity level of less than about 50 pCi/gm.

Magnesium fluoride recovery method

DOEpatents

Gay, Richard L.; McKenzie, Donald E.

1989-01-01

A method of obtaining magnesium fluoride substantially free from radioactive uranium from a slag containing the same and having a radioactivity level of at least about 7000 pCi/gm. The slag is ground to a particle size of about 200 microns or less. The ground slag is contacted with an acid under certain prescribed conditions to produce a liquid product and a particulate solid product. The particulate solid product is separated from the liquid and treated at least two more times with acid to produce a solid residue consisting essentially of magnesium fluoride substantially free of uranium and having a residual radioactivity level of less than about 1000 pCi/gm. In accordance with a particularly preferred embodiment of the invention a catalyst and an oxidizing agent are used during the acid treatment and preferably the acid is sulfuric acid having a strength of about 1.0 Normal.

Method for magnesium sulfate recovery

DOEpatents

Gay, R.L.; Grantham, L.F.

1987-08-25

A method is described for obtaining magnesium sulfate substantially free from radioactive uranium from a slag containing the same and having a radioactivity level of at least about 7,000 pCi/gm. The slag is ground to a particle size of about 200 microns or less. The ground slag is then contacted with a concentrated sulfuric acid under certain prescribed conditions to produce a liquid product and a solid product. The particulate solid product and a minor amount of the liquid is then treated to produce a solid residue consisting essentially of magnesium sulfate substantially free of uranium and having a residual radioactivity level of less than 1,000 pCi/gm. In accordance with the preferred embodiment of the invention, a catalyst and an oxidizing agent are used during the initial acid treatment and a final solid residue has a radioactivity level of less than about 50 pCi/gm.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Method for producing uranium atomic beam source

DOEpatents

Krikorian, Oscar H.

1976-06-15

A method for producing a beam of neutral uranium atoms is obtained by vaporizing uranium from a compound UM.sub.x heated to produce U vapor from an M boat or from some other suitable refractory container such as a tungsten boat, where M is a metal whose vapor pressure is negligible compared to that of uranium at the vaporization temperature. The compound, for example, may be the uranium-rhenium compound, URe.sub.2. An evaporation rate in excess of about 10 times that of conventional uranium beam sources is produced.

The influence of nepheloid layers on global model simulations of 231Pa and 230Th.

NASA Astrophysics Data System (ADS)

Basak, C.; Plancherel, Y.; Khatiwala, S.; Anderson, R. F.

2016-12-01

231Pa and 230Th in the ocean are produced at a constant ratio by Uranium decay but adsorption onto particles removes these tracers differentially. This fractionation process makes it possible to use the elemental 231Pa/230Th ratio as a paleoceanography proxy, frequently used for deriving the strength of Atlantic Meridional Overturning Circulation. The removal process, however, is further complicated by the abundance and composition of the available particle types. Understanding how dissolved tracers interact with the particle field in the ocean is key to better understand the biogeochemical cycling of these particle-reactive elements and their use as a flux tracer in present and past oceans. We here present simulations of the 231Pa/230Th ratio using the Transport Matrix Method (TMM, Khatiwala, 2007), focusing especially on the role of the nepheloid layer in controlling the distribution of these radiotracers. The model simulates each tracer separately, with advective-diffusive transport based on the ECCO ocean state estimate (Stammer et al., 2004). Sources include production by Uranium decay and dust dissolution. Radioactive decay and importantly, reversible scavenging and sedimentation are the main sinks that control the removal of the radiotracers. Similar to previous studies, we consider particle fields consisting of calcium carbonate, opal, particle organic matter, and dust. A novelty is that we explicitly consider the influence of an additional bottom particle layer (nepheloid). Simulations that include a nepheloid layer produce vertical profiles that better fit the observed distribution of 230Th and 231Pa. Specifically, observational data in the South Atlantic and eastern South Pacific indicate a mid-depth inflection (for both Pa and Th), a feature that can only be obtained if a nepheloid layer is included in the simulation. Our simulations reinforce the idea that nepheloid layers play an important role in Pa and Th cycling in the ocean (Deng et al., 2014; Hayes et al., 2015).

Physicochemical characterization of Capstone depleted uranium aerosols III: morphologic and chemical oxide analyses.

PubMed

Krupka, Kenneth M; Parkhurst, Mary Ann; Gold, Kenneth; Arey, Bruce W; Jenson, Evan D; Guilmette, Raymond A

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using x-ray diffraction (XRD), and particle morphologies were examined using scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles were spherical, occasionally with dendritic or lobed surface structures. Others appear to have fractures that perhaps resulted from abrasion and comminution, or shear bands that developed from plastic deformation of the DU material. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small bits of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.

Physicochemical Characterization of Capstone Depleted Uranium Aerosols III: Morphologic and Chemical Oxide Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krupka, Kenneth M.; Parkhurst, MaryAnn; Gold, Kenneth

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using X-ray diffraction (XRD) and particle morphologies using scanning electron microscopy/energy dispersive spectrometry (SEM/EDS). The oxidation state of a DU aerosol is importantmore » as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles appear to have been fractured (perhaps as a result of abrasion and comminution); others were spherical, occasionally with dendritic or lobed surface structures. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small chunks of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of The Journal of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.« less

High strength uranium-tungsten alloys

DOEpatents

Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

1991-01-01

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

High strength uranium-tungsten alloy process

DOEpatents

Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

1990-01-01

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

Capstone Depleted Uranium Aerosols: Generation and Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parkhurst, MaryAnn; Szrom, Fran; Guilmette, Ray

2004-10-19

In a study designed to provide an improved scientific basis for assessing possible health effects from inhaling depleted uranium (DU) aerosols, a series of DU penetrators was fired at an Abrams tank and a Bradley fighting vehicle. A robust sampling system was designed to collect aerosols in this difficult environment and continuously monitor the sampler flow rates. Aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. The resulting data provide input useful in human health risk assessments.

Release behavior of uranium in uranium mill tailings under environmental conditions.

PubMed

Liu, Bo; Peng, Tongjiang; Sun, Hongjuan; Yue, Huanjuan

2017-05-01

Uranium contamination is observed in sedimentary geochemical environments, but the geochemical and mineralogical processes that control uranium release from sediment are not fully appreciated. Identification of how sediments and water influence the release and migration of uranium is critical to improve the prevention of uranium contamination in soil and groundwater. To understand the process of uranium release and migration from uranium mill tailings under water chemistry conditions, uranium mill tailing samples from northwest China were investigated with batch leaching experiments. Results showed that water played an important role in uranium release from the tailing minerals. The uranium release was clearly influenced by contact time, liquid-solid ratio, particle size, and pH under water chemistry conditions. Longer contact time, higher liquid content, and extreme pH were all not conducive to the stabilization of uranium and accelerated the uranium release from the tailing mineral to the solution. The values of pH were found to significantly influence the extent and mechanisms of uranium release from minerals to water. Uranium release was monitored by a number of interactive processes, including dissolution of uranium-bearing minerals, uranium desorption from mineral surfaces, and formation of aqueous uranium complexes. Considering the impact of contact time, liquid-solid ratio, particle size, and pH on uranium release from uranium mill tailings, reducing the water content, decreasing the porosity of tailing dumps and controlling the pH of tailings were the key factors for prevention and management of environmental pollution in areas near uranium mines. Copyright © 2017 Elsevier Ltd. All rights reserved.

Preliminary results from the lunar prospector alpha particle spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lawson, S. L.

2001-01-01

The Lunar Prospector Alpha Particle Spectrometer (LP APS) builds on Apollo heritage and maps the distribution of outgassing sites on the Moon. The APS searches for lunar surface gas release events and maps their distribution by detecting alpha particles produced by the decay of gaseous radon-222 (5.5 MeV, 3.8 day half-life) and solid polonium-210 (5.3 MeV, 138 day half-life, but remains on the surface with a 21 year half-life as lead-210), which are radioactive daughters from the decay of uranium-238. Radon is in such small quantities that it is not released directly from the lunar interior, rather it is entrainedmore » in a stream of gases and serves as a tracer for such gases. Once released, the radon spreads out by 'bouncing' across the surface on ballistic trajectories in a random-walk process. The 3.8 day half-life of radon-222 allows the gas to spread out by several 100 km before it decays and allows the APS to detect gas release events up to a few days after they occur. The long residence time (10s of years) of the lead-210 precursor to the polonium-210 allows the mapping of gas vents which have been active over the last approximately 50 years. Because radon and polonium are daughter products of the decay of uranium, the background level of alpha particle activity is a function of the lunar crustal uranium distribution. Using radioactive radon and polonium as tracers, the Apollo 15 and 16 Command Module orbital alpha particle experiments obtained evidence for the release of gases at several sites beneath the orbit tracks, especially over the Aristarchus Plateau and Mare Fecunditatis [1]. Aristarchus crater had previously been identified by ground-based observers as the site of transient optical events [2]. The Apollo 17 surface mass spectrometer showed that argon-40 is released from the lunar interior every few months, apparently in concert with some of the shallow moonquakes that are believed to be of tectonic origin [3]. The latter tectonic events could be associated with very young scarps identified in the lunar highlands [4] and are believed to indicate continued global contraction. Such quakes could open fissures leading to the release of gases that are trapped below the surface. The detection of radon-222 outgassing events at the margins of Fecunditatis basin was surprising because the observed surface distribution of uranium and thorium do not extend sufficiently eastward to cover Fecunditatis. If the Apollo detections prove sound, then those alpha particle emissions indicate substantial subsurface concentrations of uranium-238 within Fecunditatis. A primary goal of the APS was to map gas-release events, thus allowing both an appraisal of the current level of tectonic activity on the Moon and providing a probe of subsurface uranium concentrations.« less

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-02-01

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers:more » a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).« less

Development of practical decontamination process for the removal of uranium from gravel.

PubMed

Kim, Ilgook; Kim, Gye-Nam; Kim, Seung-Soo; Choi, Jong-Won

2018-01-01

In this study, a practical decontamination process was developed to remove uranium from gravel using a soil washing method. The effects of critical parameters including particle size, H 2 SO 4 concentration, temperature, and reaction time on uranium removal were evaluated. The optimal condition for two-stage washing of gravel was found to be particle size of 1-2 mm, 1.0 M H 2 SO 4 , temperature of 60°C, and reaction time of 3 h, which satisfied the required uranium concentration for self-disposal. Furthermore, most of the extracted uranium was removed from the waste solution by precipitation, implying that the treated solution can be reused as washing solution. These results clearly demonstrated that our proposed process can be indeed a practical technique to decontaminate uranium-polluted gravel.

Process for producing enriched uranium having a .sup.235 U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

DOEpatents

Horton, James A.; Hayden, Jr., Howard W.

1995-01-01

An uranium enrichment process capable of producing an enriched uranium, having a .sup.235 U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower .sup.235 U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF.sub.6 tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a .sup.235 U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % .sup.235 U; fluorinating this enriched metallic uranium isotopic mixture to form UF.sub.6 ; processing the resultant isotopic mixture of UF.sub.6 in a gaseous diffusion process to produce a final enriched uranium product having a .sup.235 U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low .sup.235 U content UF.sub.6 having a .sup.235 U content of about 0.71 wt. % of the total uranium content of the low .sup.235 U content UF.sub.6 ; and converting this low .sup.235 U content UF.sub.6 to metallic uranium for recycle to the atomic vapor laser isotope separation process.

Process for producing enriched uranium having a {sup 235}U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

DOEpatents

Horton, J.A.; Hayden, H.W. Jr.

1995-05-30

An uranium enrichment process capable of producing an enriched uranium, having a {sup 235}U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower {sup 235}U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF{sub 6} tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a {sup 235} U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % {sup 235} U; fluorinating this enriched metallic uranium isotopic mixture to form UF{sub 6}; processing the resultant isotopic mixture of UF{sub 6} in a gaseous diffusion process to produce a final enriched uranium product having a {sup 235}U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low {sup 235}U content UF{sub 6} having a {sup 235}U content of about 0.71 wt. % of the total uranium content of the low {sup 235}U content UF{sub 6}; and converting this low {sup 235}U content UF{sub 6} to metallic uranium for recycle to the atomic vapor laser isotope separation process. 4 figs.

High strength and density tungsten-uranium alloys

DOEpatents

Sheinberg, Haskell

1993-01-01

Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

DOEpatents

Holden, R.B.; Powers, R.M.; Blaber, O.J.

1959-09-22

The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U-Mo/Al dispersion type fuel plates

NASA Astrophysics Data System (ADS)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

2014-10-01

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

Optimization of Uranium Molecular Deposition for Alpha-Counting Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Monzo, Ellen; Parsons-Moss, Tashi; Genetti, Victoria

2016-12-12

Method development for molecular deposition of uranium onto aluminum 1100 plates was conducted with custom plating cells at Lawrence Livermore National Laboratory. The method development focused primarily on variation of electrode type, which was expected to directly influence plated sample homogeneity. Solid disc platinum and mesh platinum anodes were compared and data revealed that solid disc platinum anodes produced more homogenous uranium oxide films. However, the activity distribution also depended on the orientation of the platinum electrode relative to the aluminum cathode, starting current, and material composition of the plating cell. Experiments demonstrated these variables were difficult to control undermore » the conditions available. Variation of plating parameters among a series of ten deposited plates yielded variations up to 30% in deposition efficiency. Teflon particles were observed on samples plated in Teflon cells, which poses a problem for alpha activity measurements of the plates. Preliminary electropolishing and chemical polishing studies were also conducted on the aluminum 1100 cathode plates.« less

In vitro dissolution of uranium oxide by baboon alveolar macrophages.

PubMed Central

Poncy, J L; Metivier, H; Dhilly, M; Verry, M; Masse, R

1992-01-01

In vitro cellular dissolution tests for insoluble forms of uranium oxide are technically difficult with conventional methodology using adherent alveolar macrophages. The limited number of cells per flask and the slow dissolution rate in a large volume of nutritive medium are obvious restricting factors. Macrophages in suspension cannot be substituted because they represent different and poorly reproducible functional subtypes with regard to activation and enzyme secretion. Preliminary results on the dissolution of uranium oxide using immobilized alveolar macrophages are promising because large numbers of highly functional macrophages can be cultured in a limited volume. Cells were obtained by bronchoalveolar lavages performed on baboons (Papio papio) and then immobilized after the phagocytosis of uranium octoxide (U3O8) particles in alginate beads linked with Ca2+. The dissolution rate expressed as percentage of initial uranium content in cells was 0.039 +/- 0.016%/day for particles with a count median geometric diameter of 3.84 microns(sigma g = 1.84). A 2-fold increase in the dissolution rate was observed when the same number of particles was immobilized without macrophages. These results, obtained in vitro, suggest that the U3O8 preparation investigated should be assigned to inhalation class Y as recommended by the International Commission on Radiological Protection. Future experiments are intended to clarify this preliminary work and to examine the dissolution characteristics of other particles such as uranium dioxide. It is recommended that the dissolution rate should be measured over an interval of 3 weeks, which is compatible with the survival time of immobilized cells in culture and may reveal transformation states occurring with aging of the particles. PMID:1396447

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

PROCESS FOR PRODUCING URANIUM TETRAFLUORIDE

DOEpatents

Harvey, B.G.

1954-09-14

>This patent relates to improvements in the method for producing uranium tetrafluoride by treating an aqueous solutlon of a uranyl salt at an elevated temperature with a reducing agent effective in acld solutlon in the presence of hydrofluoric acid. Uranium tetrafluoride produced this way frequentiy contains impurities in the raw material serving as the source of uranium. Uranium tetrafluoride much less contaminated with impurities than when prepared by the above method can be prepared from materials containing such impurities by first adding a small proportion of reducing agent so as to cause a small fraction, for example 1 to 5% of the uranium tetrafluoride to be precipitated, rejecting such precipitate, and then precipitating and recovering the remainder of the uranium tetrafluoride.

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments Near Church Rock, NM

DOE Office of Scientific and Technical Information (OSTI.GOV)

deLemos, J.L.; Bostick, B.C.; Quicksall, A.N.

2009-05-14

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to bemore » highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10-50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts.« less

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments near Church Rock, NM

PubMed Central

DELEMOS, JAMIE L.; BOSTICK, BENJAMIN C.; QUICKSALL, ANDREW N.; LANDIS, JOSHUA D.; GEORGE, CHRISTINE C.; SLAGOWSKI, NAOMI L.; ROCK, TOMMY; BRUGGE, DOUG; LEWIS, JOHNNYE; DURANT, JOHN L.

2008-01-01

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to be highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10–50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts. PMID:18589950

Morphological Characteristics of Particulate Material Formed from High Velocity Impact of Depleted Uranium Projectiles with Armor Targets

DTIC Science & Technology

1978-11-01

Magnification Showing Aggregation of Ultrafine Particles ; Gap Between Bars Represents 0.5 pm. .......... ... 15 iv LIST OF FIGURES (CONCLUDED) Figure Title...subsequent forma- tion of smaller particulates. An unexpected phenomenon was the formation of ultrafine particles less than 0.1 pm in diameter. These...and the highly reactive nature of pyrophoric depleted uranium. Ti ese ultrafine particles exhibited an extreme tendency to coalesce, probably due to

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-03-01

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide andmore » uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Particle detection systems and methods

DOEpatents

Morris, Christopher L.; Makela, Mark F.

2010-05-11

Techniques, apparatus and systems for detecting particles such as muons and neutrons. In one implementation, a particle detection system employs a plurality of drift cells, which can be for example sealed gas-filled drift tubes, arranged on sides of a volume to be scanned to track incoming and outgoing charged particles, such as cosmic ray-produced muons. The drift cells can include a neutron sensitive medium to enable concurrent counting of neutrons. The system can selectively detect devices or materials, such as iron, lead, gold, uranium, plutonium, and/or tungsten, occupying the volume from multiple scattering of the charged particles passing through the volume and can concurrently detect any unshielded neutron sources occupying the volume from neutrons emitted therefrom. If necessary, the drift cells can be used to also detect gamma rays. The system can be employed to inspect occupied vehicles at border crossings for nuclear threat objects.

DECONTAMINATION OF URANIUM

DOEpatents

Spedding, F.H.; Butler, T.A.

1962-05-15

A process is given for separating fission products from uranium by extracting the former into molten aluminum. Phase isolation can be accomplished by selectively hydriding the uranium at between 200 and 300 deg C and separating the hydride powder from coarse particles of fissionproduct-containing aluminum. (AEC)

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

NASA Astrophysics Data System (ADS)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Cryochemical and CVD processing of shperical carbide fuels for propulsion reactors

NASA Astrophysics Data System (ADS)

Blair, H. Thomas; Carroll, David W.; Matthews, R. Bruce

1991-01-01

Many of the nuclear propulsion reactor concepts proposed for a manned mission to Mars use a coated spherical particle fuel form similar to that used in the Rover and NERVA propulsion reactors. The formation of uranium dicarbide microspheres using a cryochemical process and the coating of the UC2 spheres with zirconium carbide using chemical vapor deposition are being developed at Los Alamos National Laboratory. The cryochemical process is described with a discussion of the variables affecting the sphere formation and carbothermic reduction to produce UC2 spheres from UO2. Emphasis is placed on minimizing the wastes produced by the process. The ability to coat particles with ZrC was recaptured, and improvements in the process and equipment were developed. Volatile organometallic precursors were investigated as alternatives to the original ZrCl4 precursor.

Combining particle-tracking and geochemical data to assess public supply well vulnerability to arsenic and uranium

USGS Publications Warehouse

Hinkle, S.R.; Kauffman, L.J.; Thomas, M.A.; Brown, C.J.; McCarthy, K.A.; Eberts, S.M.; Rosen, Michael R.; Katz, B.G.

2009-01-01

Flow-model particle-tracking results and geochemical data from seven study areas across the United States were analyzed using three statistical methods to test the hypothesis that these variables can successfully be used to assess public supply well vulnerability to arsenic and uranium. Principal components analysis indicated that arsenic and uranium concentrations were associated with particle-tracking variables that simulate time of travel and water fluxes through aquifer systems and also through specific redox and pH zones within aquifers. Time-of-travel variables are important because many geochemical reactions are kinetically limited, and geochemical zonation can account for different modes of mobilization and fate. Spearman correlation analysis established statistical significance for correlations of arsenic and uranium concentrations with variables derived using the particle-tracking routines. Correlations between uranium concentrations and particle-tracking variables were generally strongest for variables computed for distinct redox zones. Classification tree analysis on arsenic concentrations yielded a quantitative categorical model using time-of-travel variables and solid-phase-arsenic concentrations. The classification tree model accuracy on the learning data subset was 70%, and on the testing data subset, 79%, demonstrating one application in which particle-tracking variables can be used predictively in a quantitative screening-level assessment of public supply well vulnerability. Ground-water management actions that are based on avoidance of young ground water, reflecting the premise that young ground water is more vulnerable to anthropogenic contaminants than is old ground water, may inadvertently lead to increased vulnerability to natural contaminants due to the tendency for concentrations of many natural contaminants to increase with increasing ground-water residence time.

Recovery and recycling of uranium from rejected coated particles for compact high temperature reactors

NASA Astrophysics Data System (ADS)

Pai, Rajesh V.; Mollick, P. K.; Kumar, Ashok; Banerjee, J.; Radhakrishna, J.; Chakravartty, J. K.

2016-05-01

UO2 microspheres prepared by internal gelation technique were coated with pyrolytic carbon and silicon carbide using CVD technique. The particles which were not meeting the specifications were rejected. The rejected/failed UO2 based coated particles prepared by CVD technique was used for oxidation and recovery and recycling. The oxidation behaviour of sintered UO2 microspheres coated with different layers of carbon and SiC was studied by thermal techniques to develop a method for recycling and recovery of uranium from the failed/rejected coated particles. It was observed that the complete removal of outer carbon from the spheres is difficult. The crushing of microspheres enabled easier accessibility of oxygen and oxidation of carbon and uranium at 800-1000 °C. With the optimized process of multiple crushing using die & plunger and sieving the broken coated layers, we could recycle around fifty percent of the UO2 microspheres which could be directly recoated. The rest of the particles were recycled using a wet recycling method.

ALD coating of nuclear fuel actinides materials

DOEpatents

Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

2017-09-05

The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

On the distribution of uranium in hair: Non-destructive analysis using synchrotron radiation induced X-ray fluorescence microprobe techniques

NASA Astrophysics Data System (ADS)

Israelsson, A.; Eriksson, M.; Pettersson, H. B. L.

2015-06-01

In the present study the distribution of uranium in single human hair shafts has been evaluated using two synchrotron radiation (SR) based micro X-ray fluorescence techniques; SR μ-XRF and confocal SR μ-XRF. The hair shafts originated from persons that have been exposed to elevated uranium concentrations. Two different groups have been studied, i) workers at a nuclear fuel fabrication factory, exposed mainly by inhalation and ii) owners of drilled bedrock wells exposed by ingestion of water. The measurements were carried out on the FLUO beamline at the synchrotron radiation facility ANKA, Karlsruhe. The experiment was optimized to detect U with a beam size of 6.8 μm × 3 μm beam focus allowing detection down to ppb levels of U in 10 s (SR μ-XRF setup) and 70 s (SR confocal μ-XRF setup) measurements. It was found that the uranium was present in a 10-15 μm peripheral layer of the hair shafts for both groups studied. Furthermore, potential external hair contamination was studied by scanning of unwashed hair shafts from the workers. Sites of very high uranium signal were identified as particles containing uranium. Such particles, were also seen in complementary analyses using variable pressure electron microscope coupled with energy dispersive X-ray analyzer (ESEM-EDX). However, the particles were not visible in washed hair shafts. These findings can further increase the understanding of uranium excretion in hair and its potential use as a biomonitor.

Preparation of uranium fuel kernels with silicon carbide nanoparticles using the internal gelation process

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, G. W. C. M.; Lindemer, T. B.; Anderson, K. K.; Collins, J. L.

2012-08-01

The US Department of Energy continues to use the internal gelation process in its preparation of tristructural isotropic coated fuel particles. The focus of this work is to develop uranium fuel kernels with adequately dispersed silicon carbide (SiC) nanoparticles, high crush strengths, uniform particle diameter, and good sphericity. During irradiation to high burnup, the SiC in the uranium kernels will serve as getters for excess oxygen and help control the oxygen potential in order to minimize the potential for kernel migration. The hardness of SiC required modifications to the gelation system that was used to make uranium kernels. Suitable processing conditions and potential equipment changes were identified so that the SiC could be homogeneously dispersed in gel spheres. Finally, dilute hydrogen rather than argon should be used to sinter the uranium kernels with SiC.

Design of a Uranium Dioxide Spheroidization System

NASA Technical Reports Server (NTRS)

Cavender, Daniel P.; Mireles, Omar R.; Frendi, Abdelkader

2013-01-01

The plasma spheroidization system (PSS) is the first process in the development of tungsten-uranium dioxide (W-UO2) fuel cermets. The PSS process improves particle spherocity and surface morphology for coating by chemical vapor deposition (CVD) process. Angular fully dense particles melt in an argon-hydrogen plasma jet at between 32-36 kW, and become spherical due to surface tension. Surrogate CeO2 powder was used in place of UO2 for system and process parameter development. Particles range in size from 100 - 50 microns in diameter. Student s t-test and hypothesis testing of two proportions statistical methods were applied to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders show great than 800% increase in the number of spherical particles over the stock powder with the mean spherocity only mildly improved. It is recommended that powders be processed two-three times in order to reach the desired spherocity, and that process parameters be optimized for a more narrow particles size range. Keywords: spherocity, spheroidization, plasma, uranium-dioxide, cermet, nuclear, propulsion

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

Isomer Energy Source for Space Propulsion Systems

DTIC Science & Technology

2004-03-01

1,590 Engine F/W (no shield) 3.4 5.0 20.0 A similar core design replacing the fission fuel with the isomer 178Hfm2 is the starting point for this...particles interact and collide with other atoms in the fuel material, reactor core , or coolant, their energy can be transferred to thermal energy...thrust (44). The program produced several reactors that made it all the way through the testing stages of development . The reactors used uranium-235

Uranium migration in spark plasma sintered W/UO2 CERMETS

NASA Astrophysics Data System (ADS)

Tucker, Dennis S.; Wu, Yaqiao; Burns, Jatuporn

2018-03-01

W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10-15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.

The synthesis, characterization and application of iron oxide nanocrystals in magnetic separations for arsenic and uranium removal

NASA Astrophysics Data System (ADS)

Mayo, John Thomas

Arsenic and uranium in the environment are hazardous to human health and require better methods for detection and remediation. Nanocrystalline iron oxides offer a number of advantages as sorbents for water purification and environmental remediation. First, highly uniform and crystalline iron oxide nanocrystals (nMAG) were prepared using thermal decomposition of iron salts in organic solutions; for the applications of interest in this thesis, a central challenge was the adaptation of these conventional synthetic methods to the needs of low infrastructure and economically disadvantaged settings. We show here that it is possible to form highly uniform and magnetically responsive nanomaterials using starting reagents and equipment that are readily available and economical. The products of this approach, termed the 'Kitchen Synthesis', are of comparable quality and effectiveness to laboratory materials. The narrow size distributions of the iron oxides produced in the laboratory synthesis made it possible to study the size-dependence of the magnetic separation efficiency of nanocrystals; generally as the diameter of particles increased they could be removed under lower applied magnetic fields. In this work we take advantage of this size-dependence to use magnetic separation as a tool to separate broadly distributed populations of magnetic materials. Such work makes it possible to use these materials in multiplexed separation and sensing schemes. With the synthesis and magnetic separation studies of these materials completed, it was possible to optimize their applications in water purification and environmental remediation. These materials removed both uranium and arsenic from contaminated samples, and had remarkably high sorption capacities --- up to 12 wt% for arsenic and 30 wt% for uranium. The contaminated nMAG is removed from the drinking water by either retention in a sand column, filter, or by magnetic separation. The uranium adsorption process was also utilized for the enhanced detection of uranium in environmental matrices. By relying on alpha-particle detection in well-formed and dense nMAG films, it was possible to improve soil detection of uranium by more than ten-thousand-fold. Central for this work was a detailed understanding of the chemistry at the iron oxide interface, and the role of the organic coatings in mediating the sorption process.

METHOD OF OPERATING A CALUTRON

DOEpatents

Davidson, P.H.

1960-01-12

A method of operating an electromagnetic isotope separator of the calutron class is reported whereby uranium tetrachloride is produced at a controlled rate within the source rather than betng introduced therein as was formerly practiced. This is accomplished by placing a uranium-bearing material, such as uranium metal, uranium trichloride, or uranium carbide in the charge receptacle of the calutron, heating this material to about to produce uranium tetrachloride vapor at a rate controlled by the chlorine gas flow into the source. The vapor is subsequently ionized by an electric arc and mass separated by conventional calutron methods.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Far Field Modeling Methods For Characterizing Surface Detonations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garrett, A.

2015-10-08

Savannah River National Laboratory (SRNL) analyzed particle samples collected during experiments that were designed to replicate tests of nuclear weapons components that involve detonation of high explosives (HE). SRNL collected the particle samples in the HE debris cloud using innovative rocket propelled samplers. SRNL used scanning electronic microscopy to determine the elemental constituents of the particles and their size distributions. Depleted uranium composed about 7% of the particle contents. SRNL used the particle size distributions and elemental composition to perform transport calculations that indicate in many terrains and atmospheric conditions the uranium bearing particles will be transported long distances downwind.more » This research established that HE tests specific to nuclear proliferation should be detectable at long downwind distances by sampling airborne particles created by the test detonations.« less

Production of medical isotopes from a thorium target irradiated by light charged particles up to 70 MeV

NASA Astrophysics Data System (ADS)

Duchemin, C.; Guertin, A.; Haddad, F.; Michel, N.; Métivier, V.

2015-02-01

The irradiation of a thorium target by light charged particles (protons and deuterons) leads to the production of several isotopes of medical interest. Direct nuclear reaction allows the production of Protactinium-230 which decays to Uranium-230 the mother nucleus of Thorium-226, a promising isotope for alpha radionuclide therapy. The fission of Thorium-232 produces fragments of interest like Molybdenum-99, Iodine-131 and Cadmium-115g. We focus our study on the production of these isotopes, performing new cross section measurements and calculating production yields. Our new sets of data are compared with the literature and the last version of the TALYS code.

Production of medical isotopes from a thorium target irradiated by light charged particles up to 70 MeV.

PubMed

Duchemin, C; Guertin, A; Haddad, F; Michel, N; Métivier, V

2015-02-07

The irradiation of a thorium target by light charged particles (protons and deuterons) leads to the production of several isotopes of medical interest. Direct nuclear reaction allows the production of Protactinium-230 which decays to Uranium-230 the mother nucleus of Thorium-226, a promising isotope for alpha radionuclide therapy. The fission of Thorium-232 produces fragments of interest like Molybdenum-99, Iodine-131 and Cadmium-115g. We focus our study on the production of these isotopes, performing new cross section measurements and calculating production yields. Our new sets of data are compared with the literature and the last version of the TALYS code.

40 CFR Table A to Subpart D of... - Table A to Subpart D of Part 192

Code of Federal Regulations, 2013 CFR

2013-07-01

...) RADIATION PROTECTION PROGRAMS HEALTH AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Management of Uranium Byproduct Materials Pursuant to Section 84 of the Atomic... Combined radium-226 and radium-228 5 Gross alpha-particle activity (excluding radon and uranium) 15 ...

40 CFR Table A to Subpart D of... - Table A to Subpart D of Part 192

Code of Federal Regulations, 2014 CFR

2014-07-01

...) RADIATION PROTECTION PROGRAMS HEALTH AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Management of Uranium Byproduct Materials Pursuant to Section 84 of the Atomic... Combined radium-226 and radium-228 5 Gross alpha-particle activity (excluding radon and uranium) 15 ...

X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-06-01

Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batchmore » 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.« less

Evaluation of Am–Li neutron spectra data for active well type neutron multiplicity measurements of uranium

DOE PAGES

Goddard, Braden; Croft, Stephen; Lousteau, Angela; ...

2016-05-25

Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with ( α,n) neutrons from an Am-Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am-Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured spectra ofmore » various Am-Li sources to each other. This paper examines fourteen different Am-Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am-Li spectra produce similar doubles and triples count rates. Finally, the singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis.« less

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; May, Iain; Copping, Roy

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted tomore » concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.« less

Interactions between Impacting Particles and Target in Two-Phase Flow

NASA Astrophysics Data System (ADS)

Kang, Sang-Wook; Chow, Tze-Show

1996-11-01

The time-dependent interaction phenomena between a target and the incident solid particles borne by supersonic gas-jet stream have been numerically analyzed. In particular, the analysis dealt with particles such as aluminum, copper, and uranium ipinging on aluminum, copper, or uranium targets at various impact velocities ranging from 200 m/s to 1,000 m/s. Typical particle sizes were 50 to 100 micrometers. Results show considerable deformation of both the incident particles and the target when the velocity is greater than 500 m/s. Experiments performed on copper particles impacting an aluminum target demonstrate that under certain conditions (such as a supersonic gas jet issuing from a nozzle carrying solid particles) the impacts not only deform but also cause deposition of the particles on the surface. The present analysis shows the plausibility of such behavior when the particles impact the target at high velocities.

RELATIONSHIP OF URANIUM ORE DEPOSITS TO PETROLEUM AND GAS-BEARING STRUCTURES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, R.T.

eposits are located on producing or breached oil and gas structures, or in the immediate vicinity of such structures. Individual deposits associated with these structures contain ore reserves which may exceed one million tons. Data derived from a study of the known deposits should be useful in evaluating the potentiality of other areas where similar structural relations and abnormal radioactivity are known to exist. Uranium deposits located in producing oil or gas fields include a deposit of more than one million tons of uranium ore on a single salt dome in Texas, and uranium deposits in the Poison Basin, Wyoming,more » which are situated over a producing naturalgas structure, having a potential of 100,000 to 200,000 tons. Important uranium mining districts are also located near producing oil fields or near structures which may have contained oil at some time in the past. The Gas Hills district to Wyoming is on the flanks of a breached anticline and within one mile of natural-gas seeps. Deposits in the Brown's Park formation near Maybell, Colorado, are witin 10 miles of producing oil wells and natural-gas seeps are known within one mile of some of the uranium mines; and at Morrison, Colorado, uranium ore is associated with tar seeps. On th Colorado Plateau, large ore bodies with total reserves of at least 30 million tons of 0.3% U/sub 3/O/sub 8/ ore in the Ambrosia Lake district near Grants, New Mexico, and produce ore associated with asphaltite.'' The uraniferous asphaltite'' ore at Temple Mountain, Utah has been known for nearly 50 years. At both Circle Cliffs and the Inter- River area in Utah, uranium ore is associated with asphaltic material on anticlinal structures. Many other deposits are on breached strucIn Wyoming, uranium deposits in Tertiary sandstone and arkose generally lack carbon trash, but are located near oil or gas structures that contain hydrocarbons and natural gases capable of precititating uranium. Also, many uranium deposits on the Colorado Plateau have insufficient plant remains present to be the fixing agent for uranium, but petroleum and/or natural gas are proposed as possible extractants. The hydrogen sulfide contaned in natural gas or dissolved in oil-field water has been a factor in the formation of some uranium deposits. Oil-type structural traps must have been effective in localizing both petroleum and uranium ore in some districts. Although petroleum may contain small amounts of uranium, it is doubtful if either oil or natural gas are important transporting agents for uranium. Careful consideration of these various factors will provide a basis upon which to evaluate more effectively many ore producing areas. (auth)« less

PROCESS OF RECOVERING URANIUM

DOEpatents

Carter, J.M.; Larson, C.E.

1958-10-01

A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

Enhanced thermal conductivity of uranium dioxide-silicon carbide composite fuel pellets prepared by Spark Plasma Sintering (SPS)

NASA Astrophysics Data System (ADS)

Yeo, S.; Mckenna, E.; Baney, R.; Subhash, G.; Tulenko, J.

2013-02-01

Uranium dioxide (UO2)-10 vol% silicon carbide (SiC) composite fuel pellets were produced by oxidative sintering and Spark Plasma Sintering (SPS) at a range of temperatures from 1400 to 1600 °C. Both SiC whiskers and SiC powder particles were utilized. Oxidative sintering was employed over 4 h and the SPS sintering was employed only for 5 min at the highest hold temperature. It was noted that composite pellets sintered by SPS process revealed smaller grain size, reduced formation of chemical products, higher density, and enhanced interfacial contact compared to the pellets made by oxidative sintering. For given volume of SiC, the pellets with powder particles yielded a smaller grain size than pellets with SiC whiskers. Finally thermal conductivity measurements at 100 °C, 500 °C, and 900 °C revealed that SPS sintered UO2-SiC composites exhibited an increase of up to 62% in thermal conductivity compared to UO2 pellets, while the oxidative sintered composite pellets revealed significantly inferior thermal conductivity values. The current study points to the improved processing capabilities of SPS compared to oxidative sintering of UO2-SiC composites.

Radioactive equilibrium in ancient marine sediments

USGS Publications Warehouse

Breger, I.A.

1955-01-01

Radioactive equilibrium in eight marine sedimentary formations has been studied by means of direct determinations of uranium, radium and thorium. Alpha-particle counting has also been carried out in order to cross-calibrate thick-source counting techniques. The maximum deviation from radioactive equilibrium that has been noted is 11 per cent-indicating that there is probably equilibrium in all the formations analyzed. Thick-source alpha-particle counting by means of a proportional counter or an ionization chamber leads to high results when the samples contain less than about 10 p.p.m. of uranium. For samples having a higher content of uranium the results are in excellent agreement with each other and with those obtained by direct analytical techniques. The thorium contents that have been obtained correspond well to the average values reported in the literature. The uranium content of marine sediments may be appreciably higher than the average values that have been reported for sedimentary rocks. Data show that there is up to fourteen times the percentage of uranium as of thorium in the formations studied and that the percentage of thorium never exceeds that of uranium. While the proximity of a depositional environment to a land mass may influence the concentration of uranium in a marine sediment, this is not true with thorium. ?? 1955.

PROCESS FOR PREPARING URANIUM METAL

DOEpatents

Prescott, C.H. Jr.; Reynolds, F.L.

1959-01-13

A process is presented for producing oxygen-free uranium metal comprising contacting iodine vapor with crude uranium in a reaction zone maintained at 400 to 800 C to produce a vaporous mixture of UI/sub 4/ and iodine. Also disposed within the maction zone is a tungsten filament which is heated to about 1600 C. The UI/sub 4/, upon contacting the hot filament, is decomposed to molten uranium substantially free of oxygen.

Evaluation of residual uranium contamination in the dirt floor of an abandoned metal rolling mill.

PubMed

Glassford, Eric; Spitz, Henry; Lobaugh, Megan; Spitler, Grant; Succop, Paul; Rice, Carol

2013-02-01

A single, large, bulk sample of uranium-contaminated material from the dirt floor of an abandoned metal rolling mill was separated into different types and sizes of aliquots to simulate samples that would be collected during site remediation. The facility rolled approximately 11,000 tons of hot-forged ingots of uranium metal approximately 60 y ago, and it has not been used since that time. Thirty small mass (≈ 0.7 g) and 15 large mass (≈ 70 g) samples were prepared from the heterogeneously contaminated bulk material to determine how measurements of the uranium contamination vary with sample size. Aliquots of bulk material were also resuspended in an exposure chamber to produce six samples of respirable particles that were obtained using a cascade impactor. Samples of removable surface contamination were collected by wiping 100 cm of the interior surfaces of the exposure chamber with 47-mm-diameter fiber filters. Uranium contamination in each of the samples was measured directly using high-resolution gamma ray spectrometry. As expected, results for isotopic uranium (i.e., U and U) measured with the large-mass and small-mass samples are significantly different (p < 0.001), and the coefficient of variation (COV) for the small-mass samples was greater than for the large-mass samples. The uranium isotopic concentrations measured in the air and on the wipe samples were not significantly different and were also not significantly different (p > 0.05) from results for the large- or small-mass samples. Large-mass samples are more reliable for characterizing heterogeneously distributed radiological contamination than small-mass samples since they exhibit the least variation compared to the mean. Thus, samples should be sufficiently large in mass to insure that the results are truly representative of the heterogeneously distributed uranium contamination present at the facility. Monitoring exposure of workers and the public as a result of uranium contamination resuspended during site remediation should be evaluated using samples of sufficient size and type to accommodate the heterogeneous distribution of uranium in the bulk material.

Groundwater quality and the relation between pH values and occurrence of trace elements and radionuclides in water samples collected from private wells in part of the Kickapoo Tribe of Oklahoma Jurisdictional Area, central Oklahoma, 2011

USGS Publications Warehouse

Becker, Carol J.

2013-01-01

From 1999 to 2007, the Indian Health Service reported that gross alpha-particle activities and concentrations of uranium exceeded the Maximum Contaminant Levels for public drinking-water supplies in water samples from six private wells and two test wells in a rural residential neighborhood in the Kickapoo Tribe of Oklahoma Jurisdictional Area, in central Oklahoma. Residents in this rural area use groundwater from Quaternary-aged terrace deposits and the Permian-aged Garber-Wellington aquifer for domestic purposes. Uranium and other trace elements, specifically arsenic, chromium, and selenium, occur naturally in rocks composing the Garber-Wellington aquifer and in low concentrations in groundwater throughout its extent. Previous studies have shown that pH values above 8.0 from cation-exchange processes in the aquifer cause selected metals such as arsenic, chromium, selenium, and uranium to desorb (if present) from mineral surfaces and become mobile in water. On the basis of this information, the U.S. Geological Survey, in cooperation with the Kickapoo Tribe of Oklahoma, conducted a study in 2011 to describe the occurrence of selected trace elements and radionuclides in groundwater and to determine if pH could be used as a surrogate for laboratory analysis to quickly and inexpensively identify wells that might contain high concentrations of uranium and other trace elements. The pH and specific conductance of groundwater from 59 private wells were measured in the field in an area of about 18 square miles in Lincoln and Pottawatomie Counties. Twenty of the 59 wells also were sampled for dissolved concentrations of major ions, trace elements, gross alpha-particle and gross beta-particle activities, uranium, radium-226, radium-228, and radon-222 gas. Arsenic concentrations exceeded the Maximum Contaminant Level of 10 micrograms per liter in one sample having a concentration of 24.7 micrograms per liter. Selenium concentrations exceeded the Maximum Contaminant Level of 50 micrograms per liter in one sample having a concentration of 147 micrograms per liter. Both samples had alkaline pH values, 8.0 and 8.4, respectively. Uranium concentrations ranged from 0.02 to 383 micrograms per liter with 5 of 20 samples exceeding the Maximum Contaminant Level of 30 micrograms per liter; the five wells with uranium concentrations exceeding 30 micrograms per liter had pH values ranging from 8.0 to 8.5. Concentrations of uranium and radon-222 and gross alpha-particle activity showed a positive relation to pH, with the highest concentrations and activity in samples having pH values of 8.0 or above. The groundwater samples contained dissolved oxygen and high concentrations of bicarbonate; these characteristics are also factors in increasing uranium solubility. Concentrations of radium-226 and radium-228 (combined) ranged from 0.03 to 1.7 picocuries per liter, with a median concentration of 0.45 picocuries per liter for all samples. Radon-222 concentrations ranged from 95 to 3,600 picocuries per liter with a median concentration of 261 picocuries per liter. Eight samples having pH values ranging from 8.0 to 8.7 exceeded the proposed Maximum Contaminant Level of 300 picocuries per liter for radon-222. Eight samples exceeded the 15 picocuries per liter Maximum Contaminant Level for gross alpha-particle activity at 72 hours (after sample collection) and at 30 days (after the initial count); those samples had pH values ranging from 8.0 to 8.5. Gross beta-particle activity increased in 15 of 21 samples during the interval from 72 hours to 30 days. The increase in gross beta-particle activity over time probably was caused by the ingrowth and decay of uranium daughter products that emit beta particles. Water-quality data collected for this study indicate that pH values above 8.0 are associated with potentially high concentrations of uranium and radon-222 and high gross alpha-particle activity in the study area. High pH values also are associated with potentially high concentrations of arsenic, chromium, and selenium in groundwater when these elements occur in the aquifer matrix along groundwater-flow paths.

PROCESS FOR CONTINUOUSLY SEPARATING IRRADIATION PRODUCTS OF THORIUM

DOEpatents

Hatch, L.P.; Miles, F.T.; Sheehan, T.V.; Wiswall, R.H.; Heus, R.J.

1959-07-01

A method is presented for separating uranium-233 and protactinium from thorium-232 containing compositions which comprises irradiating finely divided particles of said thorium with a neutron flux to form uranium-233 and protactinium, heating the neutron-irradiated composition in a fluorine and hydrogen atmosphere to form volatile fluorides of uranium and protactinium and thereafter separating said volatile fluorides from the thorium.

77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... INFORMATION CONTACT: Glenn Tuttle, Office of Nuclear Material Safety and Safeguards, Division of Fuel Cycle...

Preparation of new nano magnetic material Fe3O4@g-C3N4 and good adsorption performance on uranium ion

NASA Astrophysics Data System (ADS)

Long, Wei; Liu, Huijun; Yan, Xueming; Fu, Li

2018-03-01

A new nano magnetic material Fe3O4@g-C3N4 was prepared by deposition reduction method, which performed good adsorption performance to uranium ion. Characterization results showed that the g-C3N4 particles were wrapped around the nano magnetic Fe3O4 particles, and the textural properties of this material was improved, so the adsorption performance to uranium ion was good. Adsorption experiments of this material demonstrated that the optimum pH value was 10, the optimum mass of adsorbent was 6.5 mg and the optimum adsorption time was 150 min in the initial concentration of 140 mg/L uranium ion solution system, and the maximum adsorption capacity was up to 352.1 mg/g and the maximum adsorption rate was more than 90%.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buck, E.C.; Cunnane, J.C.; Brown, N.R.

A combination of optical microscopy, scanning electron microscopy with backscattered electron detection (SEM/BSE), and analytical electron microscopy (AEM) is being used to determine the nature of uranium in soils from the Fernald Environmental Management Project. The information gained from these studies is being used to develop and test remediation technologies. Investigations using SEM have shown that uranium is contained within particles that are typically 1 to 100 {mu}m in diameter. Further analysis with AEM has shown that these uranium-rich regions are made up of discrete uranium-bearing phases. The distribution of these uranium phases was found to be inhomogeneous at themore » microscopic level.« less

Determination of uranium isotopic composition and 236U content of soil samples and hot particles using inductively coupled plasma mass spectrometry.

PubMed

Boulyga, S F; Becker, J S

2001-07-01

As a result of the accident at the Chernobyl nuclear power plant (NPP) the environment was contaminated with spent nuclear fuel. The 236U isotope was used in this study to monitor the spent uranium from nuclear fallout in soil samples collected in the vicinity of the Chernobyl NPP. Nuclear track radiography was applied for the identification and extraction of hot radioactive particles from soil samples. A rapid and sensitive analytical procedure was developed for uranium isotopic ratio measurement in environmental samples based on double-focusing inductively coupled plasma mass spectrometry (DF-ICP-MS) with a MicroMist nebulizer and a direct injection high-efficiency nebulizer (DIHEN). The performance of the DF-ICP-MS with a quartz DIHEN and plasma shielded torch was studied. Overall detection efficiencies of 4 x 10(-4) and 10(-3) counts per atom were achieved for 238U in DF-ICP-QMS with the MicroMist nebulizer and DIHEN, respectively. The rate of formation of uranium hydride ions UH+/U+ was 1.2 x 10(-4) and 1.4 x 10(-4), respectively. The precision of short-term measurements of uranium isotopic ratios (n = 5) in 1 microg L(-1) NBS U-020 standard solution was 0.11% (238U/235U) and 1.4% (236U/238U) using a MicroMist nebulizer and 0.25% (235U/238U) and 1.9% (236U/P38U) using a DIHEN. The isotopic composition of all investigated Chernobyl soil samples differed from those of natural uranium; i.e. in these samples the 236U/238U ratio ranged from 10(-5) to 10(-3). Results obtained with ICP-MS, alpha- and gamma-spectrometry showed differences in the migration properties of spent uranium, plutonium, and americium. The isotopic ratio of uranium was also measured in hot particles extracted from soil samples.

Deposit model for volcanogenic uranium deposits

USGS Publications Warehouse

Breit, George N.; Hall, Susan M.

2011-01-01

The International Atomic Energy Agency's tabulation of volcanogenic uranium deposits lists 100 deposits in 20 countries, with major deposits in Russia, Mongolia, and China. Collectively these deposits are estimated to contain uranium resources of approximately 500,000 tons of uranium, which amounts to 6 percent of the known global resources. Prior to the 1990s, these deposits were considered to be small (less than 10,000 tons of uranium) with relatively low to moderate grades (0.05 to 0.2 weight percent of uranium). Recent availability of information on volcanogenic uranium deposits in Asia highlighted the large resource potential of this deposit type. For example, the Streltsovskoye district in eastern Russia produced more than 100,000 tons of uranium as of 2005; with equivalent resources remaining. Known volcanogenic uranium deposits within the United States are located in Idaho, Nevada, Oregon, and Utah. These deposits produced an estimated total of 800 tons of uranium during mining from the 1950s through the 1970s and have known resources of 30,000 tons of uranium. The most recent estimate of speculative resources proposed an endowment of 200,000 tons of uranium.

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine Joyce; Stempien, John Dennis

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within amore » specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.« less

The Toxicity of Depleted Uranium

PubMed Central

Briner, Wayne

2010-01-01

Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed. PMID:20195447

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, James A.; Hayden, H. Wayne

1995-01-01

An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, J.A.; Hayden, H.W.

1995-01-10

An apparatus and method are disclosed for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode. 2 figures.

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

DOE PAGES

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; ...

2016-02-05

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

Helium interactions with alumina formed by atomic layer deposition show potential for mitigating problems with excess helium in spent nuclear fuel

NASA Astrophysics Data System (ADS)

Zhang, Shenli; Yu, Erick; Gates, Sean; Cassata, William S.; Makel, James; Thron, Andrew M.; Bartel, Christopher; Weimer, Alan W.; Faller, Roland; Stroeve, Pieter; Tringe, Joseph W.

2018-02-01

Helium gas accumulation from alpha decay during extended storage of spent fuel has potential to compromise the structural integrity the fuel. Here we report results obtained with surrogate nickel particles which suggest that alumina formed by atomic layer deposition can serve as a low volume-fraction, uniformly-distributed phase for retention of helium generated in fuel particles such as uranium oxide. Thin alumina layers may also form transport paths for helium in the fuel rod, which would otherwise be impermeable. Micron-scale nickel particles, representative of uranium oxide particles in their low helium solubility and compatibility with the alumina synthesis process, were homogeneously coated with alumina approximately 3-20 nm by particle atomic layer deposition (ALD) using a fluidized bed reactor. Particles were then loaded with helium at 800 °C in a tube furnace. Subsequent helium spectroscopy measurements showed that the alumina phase, or more likely a related nickel/alumina interface structure, retains helium at a density of at least 1017 atoms/cm3. High resolution transmission electron microscopy revealed that the thermal treatment increased the alumina thickness and generated additional porosity. Results from Monte Carlo simulations on amorphous alumina predict the helium retention concentration at room temperature could reach 1021 atoms/cm3 at 400 MPa, a pressure predicted by others to be developed in uranium oxide without an alumina secondary phase. This concentration is sufficient to eliminate bubble formation in the nuclear fuel for long-term storage scenarios, for example. Measurements by others of the diffusion coefficient in polycrystalline alumina indicate values several orders of magnitude higher than in uranium oxide, which then can also allow for helium transport out of the spent fuel.

Characterization of low concentration uranium glass working materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eppich, G. R.; Wimpenny, J. B.; Leever, M. E.

A series of uranium-doped silicate glasses were created at (Lawrence Livermore National Laboratory) LLNL, to be used as working reference material analogs for low uranium concentration research. Specifically, the aim of this effort was the generation of well-characterized glasses spanning a range of concentrations and compositions, and of sufficient homogeneity in uranium concentration and isotopic composition, for instrumentation research and development purposes. While the glasses produced here are not intended to replace or become standard materials for uranium concentration or uranium isotopic composition, it is hoped that they will help fill a current gap, providing low-level uranium glasses sufficient formore » methods development and method comparisons within the limitations of the produced glass suite. Glasses are available for research use by request.« less

PREPARATION OF METAL POWDER COMPACTS PRIOR TO PRESSING

DOEpatents

Mansfield, H.

1958-08-26

A method of fabricating uranium by a powder metallurgical technique is described. It consists in introducing powdered uranium hydride into a receptacle shaped to coincide with the coatour of the die cavity and heating the hydride so that it decomposes to uranium metal. The metal particles cohere in the shapw of the receptacle and thereafter the prefurmed metal powder is pressed and sintered to obtain a dense compact.

Properties, use and health effects of depleted uranium (DU): a general overview.

PubMed

Bleise, A; Danesi, P R; Burkart, W

2003-01-01

Depleted uranium (DU), a waste product of uranium enrichment, has several civilian and military applications. It was used as armor-piercing ammunition in international military conflicts and was claimed to contribute to health problems, known as the Gulf War Syndrome and recently as the Balkan Syndrome. This led to renewed efforts to assess the environmental consequences and the health impact of the use of DU. The radiological and chemical properties of DU can be compared to those of natural uranium, which is ubiquitously present in soil at a typical concentration of 3 mg/kg. Natural uranium has the same chemotoxicity, but its radiotoxicity is 60% higher. Due to the low specific radioactivity and the dominance of alpha-radiation no acute risk is attributed to external exposure to DU. The major risk is DU dust, generated when DU ammunition hits hard targets. Depending on aerosol speciation, inhalation may lead to a protracted exposure of the lung and other organs. After deposition on the ground, resuspension can take place if the DU containing particle size is sufficiently small. However, transfer to drinking water or locally produced food has little potential to lead to significant exposures to DU. Since poor solubility of uranium compounds and lack of information on speciation precludes the use of radioecological models for exposure assessment, biomonitoring has to be used for assessing exposed persons. Urine, feces, hair and nails record recent exposures to DU. With the exception of crews of military vehicles having been hit by DU penetrators, no body burdens above the range of values for natural uranium have been found. Therefore, observable health effects are not expected and residual cancer risk estimates have to be based on theoretical considerations. They appear to be very minor for all post-conflict situations, i.e. a fraction of those expected from natural radiation.

Determination of uranium in clinical and environmental samples by FIAS-ICPMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karpas, Z.; Lorber, A.; Halicz, L.

Uranium may enter the human body through ingestion or inhalation. Ingestion of uranium compounds through the diet, mainly drinking water, is a common occurrence, as these compounds are present in the biosphere. Inhalation of uranium-containing particles is mainly an occupational safety problem, but may also take place in areas where uranium compounds are abundant. The uranium concentration in urine samples may serve as an indication of the total uranium body content. A method based on flow injection and inductively coupled plasma mass spectrometry (FIAS-ICPMS) was found to be most suitable for determination of uranium in clinical samples (urine and serum),more » environmental samples (seawater, wells and carbonate rocks) and in liquids consumed by humans (drinking water and commercial beverages). Some examples of the application of the FIAS-ICPMS method are reviewed and presented here.« less

Swelling Mechanisms of UO2 Lattices with Defect Ingrowths

PubMed Central

Günay, Seçkin D.

2015-01-01

The swelling that occurs in uranium dioxide as a result of radiation-induced defect ingrowth is not fully understood. Experimental and theoretical groups have attempted to explain this phenomenon with various complex theories. In this study, experimental lattice expansion and lattice super saturation were accurately reproduced using a molecular dynamics simulation method. Based on their resemblance to experimental data, the simulation results presented here show that fission induces only oxygen Frenkel pairs while alpha particle irradiation results in both oxygen and uranium Frenkel pair defects. Moreover, in this work, defects are divided into two sub-groups, obstruction type defects and distortion type defects. It is shown that obstruction type Frenkel pairs are responsible for both fission- and alpha-particle-induced lattice swelling. Relative lattice expansion was found to vary linearly with the number of obstruction type uranium Frenkel defects. Additionally, at high concentrations, some of the obstruction type uranium Frenkel pairs formed diatomic and triatomic structures with oxygen ions in their octahedral cages, increasing the slope of the linear dependence. PMID:26244777

Preparation of carbon coated Fe3O4 nanoparticles for magnetic separation of uranium

NASA Astrophysics Data System (ADS)

Zhang, Xiaofei; Wang, Jun

2018-01-01

Uranium(VI) was removed from aqueous solutions using carbon coated Fe3O4 nanoparticles (Fe3O4@C). Batch experiments were conducted to study the effects of initial pH, shaking time and temperature on uranium sorption efficiency. It was found that the maximum adsorption capacity of the Fe3O4@C toward uranium(VI) was ∼120.20 mg g-1 when the initial uranium(VI) concentration was 100 mg L-1, displaying a high efficiency for the removal of uranium(VI) ions. Kinetics of the uranium(VI) removal is found to follow pseudo-second-order rate equation. In addition, the uranium(VI)-loaded Fe3O4@C nanoparticles can be recovered easily from aqueous solution by magnetic separation and regenerated by acid treatment. Present study suggested that magnetic Fe3O4@C composite particles can be used as an effective and recyclable adsorbent for the removal of uranium(VI) from aqueous solutions.

Acceptance Test Data for BWXT Coated Particle Batches 93172B and 93173B—Defective IPyC and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Helmreich, Grant W.; Dyer, John A.

Coated particle batches J52O-16-93172B and J52O-16-93173B were produced by Babcock and Wilcox Technologies (BWXT) as part of the production campaign for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), but were not used in the final fuel composite. However, these batches may be used as demonstration production-scale coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture ofmore » 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93172A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017b]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93172B).« less

PRODUCTION OF URANIUM MONOCARBIDE

DOEpatents

Powers, R.M.

1962-07-24

A method of making essentially stoichiometric uranium monocarbide by pelletizing a mixture of uranium tetrafluoride, silicon, and carbon and reacting the mixture at a temperature of approximately 1500 to 1700 deg C until the reaction goes to completion, forming uranium monocarbide powder and volatile silicon tetrafluoride, is described. The powder is then melted to produce uranium monocarbide in massive form. (AEC)

Method of Making Uranium Dioxide Bodies

DOEpatents

Wilhelm, H. A.; McClusky, J. K.

1973-09-25

Sintered uranium dioxide bodies having controlled density are produced from U.sub.3 O.sub.8 and carbon by varying the mole ratio of carbon to U.sub.3 O.sub.8 in the mixture, which is compressed and sintered in a neutral or slightly oxidizing atmosphere to form dense slightly hyperstoichiometric uranium dioxide bodies. If the bodies are to be used as nuclear reactor fuel, they are subsequently heated in a hydrogen atmosphere to achieve stoichiometry. This method can also be used to produce fuel elements of uranium dioxide -- plutonium dioxide having controlled density.

PROCESS OF RECOVERING URANIUM

DOEpatents

Kilner, S.B.

1959-12-29

A method is presented for separating and recovering uranium from a complex mixure of impurities. The uranium is dissolved to produce an aqueous acidic solution including various impurities. In accordance with one method, with the uranium in the uranyl state, hydrogen cyanide is introduced into the solution to complex the impurities. Subsequently, ammonia is added to the solution to precipitate the uraniunn as ammonium diuranate away from the impurities in the solution. Alternatively, the uranium is precipitated by adding an alkaline metal hydroxide. In accordance with the second method, the uranium is reduced to the uranous state in the solution. The reduced solution is then treated with solid alkali metal cyanide sufficient to render the solution about 0.1 to 1.0 N in cyanide ions whereat cyanide complex ions of the metal impurities are produced and the uranium is simultaneously precipituted as uranous hydroxide. Alternatively, hydrogen cyanide may be added to the reduced solution and the uranium precipitated subsequently by adding ammonium hydroxide or an alkali metal hydroxide. Other refinements of the method are also disclosed.

Production of near-full density uranium nitride microspheres with a hot isostatic press

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMurray, Jacob W.; Kiggans, Jr., Jim O.; Helmreich, Grant W.

Depleted uranium nitride (UN) kernels with diameters ranging from 420 to 858 microns and theoretical densities (TD) between 87 and 91 percent were postprocessed using a hot isostatic press (HIP) in an argon gas media. This treatment was shown to increase the TD up to above 97%. Uranium nitride is highly reactive with oxygen. Therefore, a novel crucible design was implemented to remove impurities in the argon gas via in situ gettering to avoid oxidation of the UN kernels. The density before and after each HIP procedure was calculated from average weight, volume, and ellipticity determined with established characterization techniquesmore » for particle. Furthermore, micrographs confirmed the nearly full densification of the particles using the gettering approach and HIP processing parameters investigated in this work.« less

Uranium comminution age tested by the eolian deposits on the Chinese Loess Plateau

NASA Astrophysics Data System (ADS)

Li, Le; Liu, Xiangjun; Li, Tao; Li, Laifeng; Zhao, Liang; Ji, Junfeng; Chen, Jun; Li, Gaojun

2017-06-01

The 234U/238U ratio of fine particles can record the time since their separation from bed rock because of the disruption of uranium series equilibrium introduced by the recoil of daughter 234Th nuclei (precursor of 234U) out of particle surfaces during the decay of 238U. Application of the uranium comminution age method, which has great potential in tracing production and transportation of sediments is however complicated by the weathering dissolution of 234U depleted particle surfaces, the difficulty in determining the fraction of recoiled nuclei, and the precipitation of exogenetic 234U. Here we minimize these complications by using a newly developed precise size separation using electroformed sieve, and a chemical protocol that involves reductive and oxidative leaching. Eolian deposits collected from the Chinese Loess Plateau (CLP) were used to test the validity of our method. Possible effects of weathering dissolution were also evaluated by comparing samples with different weathering intensities. The results show decreasing 234U/238U ratios in fine eolian particles with increasing sedimentation age, agreeing well with the theoretical prediction of the comminution age model. This successful application of the uranium comminution age approach to the eolian deposits on the CLP is also aided by a stable dust source, the low weathering intensity, the lack of consolidation, and the well-defined age model of the deposits. A transportation time of 242 ± 18 ka was calculated for the eolian deposits, which indicates a long residence time, and thus extensive mixing, of the dust particles in source regions, partly explaining the stable and homogeneous composition of the eolian dust over glacial-interglacial cycles.

Developing uranium dicarbide-graphite porous materials for the SPES project

NASA Astrophysics Data System (ADS)

Biasetto, L.; Zanonato, P.; Carturan, S.; Di Bernardo, P.; Colombo, P.; Andrighetto, A.; Prete, G.

2010-09-01

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO 2, U 3O 8 and UO 3), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U 3O 8 and UO 3 oxides to UO 2 before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC 2 and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.

Combined effects of alpha particles and depleted uranium on Zebrafish (Danio rerio) embryos

PubMed Central

Ng, Candy Y.P.; Pereira, Sandrine; Cheng, Shuk Han; Adam-Guillermin, Christelle; Garnier-Laplace, Jacqueline; Yu, Kwan Ngok

2016-01-01

The combined effects of low-dose or high-dose alpha particles and depleted uranium (DU) in Zebrafish (Danio rerio) embryos were studied. Three schemes were examined—(i) [ILUL]: 0.44 mGy alpha-particle dose + 10 µg/l DU exposure, (ii) [IHUH]: 4.4 mGy alpha-particle dose + 100 µg/l DU exposure and (iii) [IHUL]: 4.4 mGy alpha-particle dose + 10 µg/l DU exposure—in which Zebrafish embryos were irradiated with alpha particles at 5 h post fertilization (hpf) and/or exposed to uranium at 5–6 hpf. The results were also compared with our previous work, which studied the effects of [ILUH]: 0.44 mGy alpha-particle dose + 100 µg/l DU exposure. When the Zebrafish embryos developed to 24 hpf, the apoptotic signals in the entire embryos, used as the biological endpoint for this study, were quantified. Our results showed that [ILUL] and [IHUL] led to antagonistic effects, whereas [IHUH] led to an additive effect. The effect found for the previously studied case of [ILUH] was difficult to define because it was synergistic with reference to the 100 µg/l DU exposure, but it was antagonistic with reference to the 0.44 mGy alpha-particle dose. All the findings regarding the four different schemes showed that the combined effects critically depended on the dose response to each individual stressor. We also qualitatively explained these findings in terms of promotion of early death of cells predisposed to spontaneous transformation by alpha particles, interacting with the delay in cell death resulting from various concentrations of DU exposure. PMID:26937024

PRODUCTION OF URANIUM TETRAFLUORIDE

DOEpatents

Shaw, W.E.; Spenceley, R.M.; Teetzel, F.M.

1959-08-01

A method is presented for producing uranium tetrafluoride from the gaseous hexafluoride by feeding the hexafluoride into a high temperature zone obtained by the recombination of molecularly dissociated hydrogen. The molal ratio of hydrogen to uranium hexnfluoride is preferably about 3 to 1. Uranium tetrafluoride is obtained in a finely divided, anhydrous state.

Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contractmore » to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.« less

On the reactive occlusion of the (uranium trichloride + lithium chloride + potassium chloride) eutectic salt in zeolite 4A

NASA Astrophysics Data System (ADS)

Lexa, Dusan; Leibowitz, Leonard; Kropf, Jeremy

2000-03-01

The interaction between the (uranium trichloride + lithium chloride + potassium chloride) eutectic salt and zeolite 4A has been studied by temperature-resolved synchrotron powder X-ray diffraction, evolved gas analysis and differential scanning calorimetry, between 300 and 900 K. The onset of salt occlusion by the zeolite has been detected at 450 K. Evidence of a reaction between zeolitic water and uranium trichloride, leading to the formation of uranium dioxide, has appeared at 600 K. The uranium dioxide particle size increases from 2 nm at 600 K to 25 nm at 900 K - an indication of their extra-zeolitic location. No appreciable degradation of the zeolite structure has been observed.

Depleted uranium dust from fired munitions: physical, chemical and biological properties.

PubMed

Mitchel, R E J; Sunder, S

2004-07-01

This paper reports physical, chemical and biological analyses of samples of dust resulting from munitions containing depleted uranium (DU) that had been live-fired and had impacted an armored target. Mass spectroscopic analysis indicated that the average atom% of U was 0.198 +/- 0.10, consistent with depleted uranium. Other major elements present were iron, aluminum, and silicon. About 47% of the total mass was particles with diameters <300 microm, of which about 14% was <10 microm. X-ray diffraction analysis indicated that the uranium was present in the sample as uranium oxides-mainly U3O7 (47%), U3O8 (44%) and UO2 (9%). Depleted uranium dust, instilled into the lungs or implanted into the muscle of rats, contained a rapidly soluble uranium component and a more slowly soluble uranium component. The fraction that underwent dissolution in 7 d declined exponentially with increasing initial burden. At the lower lung burdens tested (<15 microg DU dust/lung) about 14% of the uranium appeared in urine within 7 d. At the higher lung burdens tested (~80-200 microg DU dust/lung) about 5% of the DU appeared in urine within 7 d. In both cases about 50% of that total appeared in urine within the first day. DU implanted in muscle similarly showed that about half of the total excreted within 7 d appeared in the first day. At the lower muscle burdens tested (<15 microg DU dust/injection site) about 9% was solubilized within 7 d. At muscle burdens >35 microg DU dust/injection site about 2% appeared in urine within 7 d. Natural uranium (NU) ore dust was instilled into rat lungs for comparison. The fraction dissolving in lung showed a pattern of exponential decline with increasing initial burden similar to DU. However, the decline was less steep, with about 14% appearing in urine for lung burdens up to about 200 microg NU dust/lung and 5% at lung burdens >1,100 microg NU dust/lung. NU also showed both a fast and a more slowly dissolving component. At the higher lung burdens of both DU and NU that showed lowered urine excretion rates, histological evidence of kidney damage was seen. Kidney damage was not seen with the muscle burdens tested. DU dust produced kidney damage at lower lung burdens and lower urine uranium levels than NU dust, suggesting that other toxic metals in DU dust may contribute to the damage.

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2010 CFR

2010-04-01

...) Application of 50 percent test. The 50 percent test described in subparagraph (1) of this paragraph is applied... uranium concentrates (known in the industry as “yellow cake”), and nuclear fuel materials derived from the refining of uranium ore and uranium concentrates, or produced in a nuclear reaction, including— (a) Uranium...

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2011 CFR

2011-04-01

...) Application of 50 percent test. The 50 percent test described in subparagraph (1) of this paragraph is applied... uranium concentrates (known in the industry as “yellow cake”), and nuclear fuel materials derived from the refining of uranium ore and uranium concentrates, or produced in a nuclear reaction, including— (a) Uranium...

Synchrotron X-ray characterization of mackinawite and uraninite relevant to bio-remediation of groundwater contaminated with uranium

NASA Astrophysics Data System (ADS)

Carpenter, J.; Hyun, S.; Hayes, K. F.

2010-12-01

Uranium (U) originating from mining operations for weapon manufacturing and nuclear energy production is a significant radionuclide contaminant in groundwater local to uranium mining, uranium milling, and uranium mill tailing (UMT) storage sites. In the USA, the Department of Energy (DOE) is currently overseeing approximately 24 Uranium Mill Tailing Remediation Action (UMTRA) sites which have collectively processed over 27 million tons of uranium ore1,2. In-Situ microbial bio-reduction of the highly mobile U6+ ion into the dramatically less mobile U4+ ion has been demonstrated as an effective remedial process to inhibit uranium migration in the aqueous phase3. The resistance of this process to oxidization and possible remobilization of U when bioremediation stops (and oxidants such as oxygen from the air or nitrate in water diffuse into the formation) in the long term is not known. UMTRA site studies3 have shown that iron sulfide solids are produced by sulfate reducing bacteria (SRB) during U bioremediation, and some forms of these iron sulfide solids are known to be effective oxidant scavengers, potentially protecting against re-oxidation and thus remobilization of U. This work is investigating the role of iron sulfide solids in the long-term immobilization of reduced U compounds after bioremediation is completed in groundwater local to UMTRA sites. Re-oxidation tests are being performed in packed media columns loaded with both FeS and U solids. High quality mackinawite (FeS), and uraninite (UO2) have been synthesized in our laboratory via a wet chemistry approach. These synthetic materials are expected to mimic the naturally occurring and biogenic materials present in biologically stimulated UMTRA sites. In order to establish the initial conditions of the prepared experimental columns and to compare synthetic and biogenic FeS and UO2, these synthesized materials have been characterized with synchrotron radiation at the Stanford Synchrotron Radiation Lightsource using synchrotron x-ray powder diffraction (SXRD) and extended x-ray absorption fine structure (EXAFS). SXRD data were collected and analyzed with profile fitting to determine lattice parameters and crystallite size for comparison with published values for both biogenic and synthetic materials. This is particularly of interest for UO2, as there is very little information on particle size and lattice parameters for synthetic UO2 in the literature. Profile fitting of the SXRD data for FeS gives lattice parameters of a = b = 3.668 and a mean crystallite size of 5 to 8 nm. Both of these values are in good agreement with published values. For fresh UO2, lattice parameters were determined as a = b = c = 5.4 nm for both freshly synthesized and aged (3 months) UO2 and particle size was determined to be 3.5 nm for fresh UO2 and 5.83 nm for aged UO2. This suggests a growth mechanism for crystallites over time, and an inferred decrease in reactivity.

Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

NASA Astrophysics Data System (ADS)

Goddard, Braden

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

REFRACTORY ARTICLE AND PROCESS OF MANUFACTURING SAME

DOEpatents

Hamilton, N.E.

1957-12-10

A method is described for fabricating improved uranium oxide crucibles. In the past, such crucibles have lacked mechanical strength due to the poor cohesion of the uranium oxide particles. This difficulty has now been overcome by admixing with the uranium oxide a quantity of a refractory oxide binder, and dry pressing and sintering the resulting mixture into the desired shape. Suitable as binders are BeO, CaO, Al/sub 2/C/sub 3/, and ThO/sub 2/ among others.

Evaluation of Settler Tank Thermal Stability during Solidification and Disposition to ERDF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stephenson, David E.; Delegard, Calvin H.; Schmidt, Andrew J.

2015-03-30

Ten 16-foot-long and 20-inch diameter horizontal tanks currently reside in a stacked 2×5 (high) array in the ~20,000-gallon water-filled Weasel Pit of the 105-KW Fuel Storage Basin on the US-DOE Hanford Site. These ten tanks are part of the Integrated Water Treatment System used to manage water quality in the KW Basin and are called “settler” tanks because of their application in removing particles from the KW Basin waters. Based on process knowledge, the settler tanks are estimated to contain about 124 kilograms of finely divided uranium metal, 22 kg of uranium dioxide, and another 55 kg of other radioactivemore » sludge. The Sludge Treatment Project (STP), managed by CH2MHill Plateau Remediation Company (CHPRC) is charged with managing the settler tanks and arranging for their ultimate disposal by burial in ERDF. The presence of finely divided uranium metal in the sludge is of concern because of the potential for thermal runaway reaction of the uranium metal with water and the formation of flammable hydrogen gas as a product of the uranium-water reaction. Thermal runaway can be instigated by external heating. The STP commissioned a formal Decision Support Board (DSB) to consider options and provide recommendations to manage and dispose of the settler tanks and their contents. Decision criteria included consideration of the project schedule and longer-term deactivation, decontamination, decommissioning, and demolition (D4) of the KW Basin. The DSB compared the alternatives and recommended in-situ grouting, size-reduction, and ERDF disposal as the best of six candidate options for settler tank treatment and disposal. It is important to note that most grouts contain a complement of Portland cement as the binding agent and that Portland cement curing reactions generate heat. Therefore, concern is raised that the grouting of the settler tank contents may produce heating sufficient to instigate thermal runaway reactions in the contained uranium metal sludge.« less

The D-D Neutron Generator as an Alternative to Am(Li) Isotopic Neutron Source in the Active Well Coincidence Counter

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, Robert Dennis; Cleveland, Steven L.

The 235U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques due to the limited penetration of the item by the characteristic 235U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Collar (UNCL) and 252Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., 252Cf or Am(Li)] to achieve better measurement accuracies than are possible usingmore » gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The 252Cf shuffler systems rely on the use of high-output 252Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every five to seven years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. In this work, we made use of Oak Ridge National Laboratory’s (ORNL’s) Large Volume Active Well Coincidence Counter (LV-AWCC) and a commercially available deuterium-deuterium (D-D) neutron generator to examine the potential of the D-D neutron generator as an alternative to the isotopic sources. We present the performance of the LV-AWCC with D-D generator for the assay of 235U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium (DU), low enriched uranium (LEU), and highly enriched uranium (HEU) items.« less

Distribution of Pd, Ag & U in the SiC Layer of an Irradiated TRISO Fuel Particle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas M. Lillo; Isabella J. van Rooyen

2014-08-01

The distribution of silver, uranium and palladium in the silicon carbide (SiC) layer of an irradiated TRISO fuel particle was studied using samples extracted from the SiC layer using focused ion beam (FIB) techniques. Transmission electron microscopy in conjunction with energy dispersive x-ray spectroscopy was used to identify the presence of the specific elements of interest at grain boundaries, triple junctions and precipitates in the interior of SiC grains. Details on sample fabrication, errors associated with measurements of elemental migration distances and the distances migrated by silver, palladium and uranium in the SiC layer of an irradiated TRISO particle frommore » the AGR-1 program are reported.« less

PRODUCTION OF URANIUM

DOEpatents

Spedding, F.H.; Wilhelm, H.A.; Keller, W.H.

1958-04-15

The production of uranium metal by the reduction of uranium tetrafluoride is described. Massive uranium metal of high purily is produced by reacting uranium tetrafluoride with 2 to 20% stoichiometric excess of magnesium at a temperature sufficient to promote the reaction and then mantaining the reaction mass in a sealed vessel at temperature in the range of 1150 to 2000 d C, under a superatomospheric pressure of magnesium for a period of time sufficient 10 allow separation of liquid uranium and liquid magnesium fluoride into separate layers.

Method for fabricating uranium foils and uranium alloy foils

DOEpatents

Hofman, Gerard L [Downers Grove, IL; Meyer, Mitchell K [Idaho Falls, ID; Knighton, Gaven C [Moore, ID; Clark, Curtis R [Idaho Falls, ID

2006-09-05

A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

Measuring Aerosols Generated Inside Armoured Vehicles Perforated by Depleted Uranium Ammunition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parkhurst, MaryAnn

2003-01-01

In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the U.S. Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects of soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterized. A robust sampling system was designed to collect aerosols in this difficult environment and to monitor continuously the sampler flowmore » rates. Interior aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. These data will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.« less

Rapid fusion method for the determination of refractory thorium and uranium isotopes in soil samples

DOE PAGES

Maxwell, Sherrod L.; Hutchison, Jay B.; McAlister, Daniel R.

2015-02-14

Recently, approximately 80% of participating laboratories failed to accurately determine uranium isotopes in soil samples in the U.S Department of Energy Mixed Analyte Performance Evaluation Program (MAPEP) Session 30, due to incomplete dissolution of refractory particles in the samples. Failing laboratories employed acid dissolution methods, including hydrofluoric acid, to recover uranium from the soil matrix. The failures illustrate the importance of rugged soil dissolution methods for the accurate measurement of analytes in the sample matrix. A new rapid fusion method has been developed by the Savannah River National Laboratory (SRNL) to prepare 1-2 g soil sample aliquots very quickly, withmore » total dissolution of refractory particles. Soil samples are fused with sodium hydroxide at 600 ºC in zirconium crucibles to enable complete dissolution of the sample. Uranium and thorium are separated on stacked TEVA and TRU extraction chromatographic resin cartridges, prior to isotopic measurements by alpha spectrometry on cerium fluoride microprecipitation sources. Plutonium can also be separated and measured using this method. Batches of 12 samples can be prepared for measurement in <5 hours.« less

Formation and Geological Sequestration of Uranium Nanoparticles in Deep Granitic Aquifer

PubMed Central

Suzuki, Yohey; Mukai, Hiroki; Ishimura, Toyoho; Yokoyama, Takaomi D.; Sakata, Shuhei; Hirata, Takafumi; Iwatsuki, Teruki; Mizuno, Takashi

2016-01-01

The stimulation of bacterial activities that convert hexavalent uranium, U(VI), to tetravalent uranium, U(IV), appears to be feasible for cost-effective remediation of contaminated aquifers. However, U(VI) reduction typically results in the precipitation of U(IV) particles less than 5 nanometers in diameter, except for environmental conditions enriched with iron. Because these tiny particles are mobile and susceptible to oxidative dissolution after the termination of nutrient injection, in situ bioremediation remains to be impractical. Here we show that U(IV) nanoparticles of coffinite (U(SiO4)1−x(OH)4x) formed in fracture-filling calcium carbonate in a granitic aquifer. In situ U-Pb isotope dating demonstrates that U(IV) nanoparticles have been sequestered in the calcium carbonate for at least 1 million years. As the microbiologically induced precipitation of calcium carbonate in aquifer systems worldwide is extremely common, we anticipate simultaneous stimulation of microbial activities for precipitation reactions of calcium carbonate and U(IV) nanoparticles, which leads to long-term sequestration of uranium and other radionuclides in contaminated aquifers and deep geological repositories. PMID:26948389

Formation and Geological Sequestration of Uranium Nanoparticles in Deep Granitic Aquifer.

PubMed

Suzuki, Yohey; Mukai, Hiroki; Ishimura, Toyoho; Yokoyama, Takaomi D; Sakata, Shuhei; Hirata, Takafumi; Iwatsuki, Teruki; Mizuno, Takashi

2016-03-07

The stimulation of bacterial activities that convert hexavalent uranium, U(VI), to tetravalent uranium, U(IV), appears to be feasible for cost-effective remediation of contaminated aquifers. However, U(VI) reduction typically results in the precipitation of U(IV) particles less than 5 nanometers in diameter, except for environmental conditions enriched with iron. Because these tiny particles are mobile and susceptible to oxidative dissolution after the termination of nutrient injection, in situ bioremediation remains to be impractical. Here we show that U(IV) nanoparticles of coffinite (U(SiO4)1-x(OH)4x) formed in fracture-filling calcium carbonate in a granitic aquifer. In situ U-Pb isotope dating demonstrates that U(IV) nanoparticles have been sequestered in the calcium carbonate for at least 1 million years. As the microbiologically induced precipitation of calcium carbonate in aquifer systems worldwide is extremely common, we anticipate simultaneous stimulation of microbial activities for precipitation reactions of calcium carbonate and U(IV) nanoparticles, which leads to long-term sequestration of uranium and other radionuclides in contaminated aquifers and deep geological repositories.

Gamma-ray spectroscopy of 131Sn81 via the (9Be, 8Be γ) reaction

NASA Astrophysics Data System (ADS)

Burcher, Sean; Bey, A.; Jones, K.; Ahn, S. H.; Ayres, A.; Schmitt, K. T.; Allmond, J.; Galindo-Urribari, A.; Radford, D. C.; Liang, J. F.; Neseraja, C. D.; Pain, S. D.; Pittman, S. T.; Smith, M. S.; Stracener, D. W.; Varner, R. L.; Bardayan, D. W.; O'Malley, P. D.; Cizewski, J. A.; Howard, M. E.; Manning, B. M.; Garcia Ruiz, R. F.; Kozub, R. L.; Matos, M.; Padilla-Rodal, E.

2016-09-01

Nuclear data in the region of the doubly-magic nucleus 132Sn82 is useful for benchmarking nuclear structure theories due to the clean single-particle nature of the nuclear wavefunction near the closed shells. At the Holifield Radioactive Ion Beam Facility (HRIBF) neutron-rich beams in the 132Sn82 region were produced via proton-induced fission of a Uranium-Carbide target. The CLARION array of HPGe detectors was coupled with the HyBall array of CsI detectors to allow for particle-gamma coincidence measurements. The gamma-ray de-excitation of the four lowest lying single-neutron states has been observed for the first time via the (9Be,8Be γ) reaction. The excitation energy of these states have been measured to higher precision than was possible with the previous charged particle measurement. This work was supported in part by the U.S. Department of Energy and the National Science Foundation.

Uranium in Surface Waters and Sediments Affected by Historical Mining in the Denver West 1:100,000 Quadrangle, Colorado

USGS Publications Warehouse

Zielinski, Robert A.; Otton, James K.; Schumann, R. Randall; Wirt, Laurie

2008-01-01

Geochemical sampling of 82 stream waters and 87 stream sediments within mountainous areas immediately west of Denver, Colorado, was conducted by the U.S. Geological Survey in October 1994. The primary purpose was to evaluate regionally the effects of geology and past mining on the concentration and distribution of uranium. The study area contains uranium- and thorium-rich bedrock, numerous noneconomic occurrences of uranium minerals, and several uranium deposits of variable size and production history. During the sampling period, local streams had low discharge and were more susceptible to uranium-bearing acid drainage originating from historical mines of base- and precious-metal sulfides. Results indicated that the spatial distribution of Precambrian granites and metamorphic rocks strongly influences the concentration of uranium in stream sediments. Within-stream transport increases the dispersion of uranium- and thorium rich mineral grains derived primarily from granitic source rocks. Dissolved uranium occurs predominantly as uranyl carbonate complexes, and concentrations ranged from less than 1 to 65 micrograms per liter. Most values were less than 5 micrograms per liter, which is less than the current drinking water standard of 30 micrograms per liter and much less than locally applied aquatic-life toxicity standards of several hundred micrograms per liter. In local streams that are affected by uranium-bearing acid mine drainage, dissolved uranium is moderated by dilution and sorptive uptake by stream sediments. Sorbents include mineral alteration products and chemical precipitates of iron- and aluminum-oxyhydroxides, which form where acid drainage enters streams and is neutralized. Suspended uranium is relatively abundant in some stream segments affected by nearby acid drainage, which likely represents mobilization of these chemical precipitates. The 234U/238U activity ratio of acid drainage (0.95-1.0) is distinct from that of local surface waters (more than 1.05), and this distinctive isotopic composition may be preserved in iron-oxyhydroxide precipitates of acid drainage origin. The study area includes a particularly large vein-type uranium deposit (Schwartzwalder mine) with past uranium production. Stream water and sediment collected downstream from the mine's surface operations have locally anomalous concentrations of uranium. Fine-grained sediments downstream from the mine contain rare minute particles (10-20 micrometers) of uraninite, which is unstable in a stream environment and thus probably of recent origin related to mining. Additional rare particles of very fine grained (less than 5 micrometer) barite likely entered the stream as discharge from settling ponds in which barite precipitation was formerly used to scavenge dissolved radium from mine effluent.

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

Laser and gas centrifuge enrichment

NASA Astrophysics Data System (ADS)

Heinonen, Olli

2014-05-01

Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

PROCESS FOR UTILIZING ORGANIC ORTHOPHOSPHATE EXTRACTANTS

DOEpatents

Grinstead, R.R.

1958-11-11

A process is presented for recovering uranium from its ores, the steps comprising producing the uranium in solution in the trivalent state, extracting the uranium from solution in an lmmiscible organic solvent extract phase which lncludes mono and dialkyl orthophosphorlc acid esters having a varying number of carbon atoms on the alkyl substituent, amd recovering the uranium from tbe extract phase.

IMPROVEMENTS IN OR RELATING TO THE PRODUCTION OF SINTERED URANIUM DIOXIDE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, L.E.; Harrison, J.D.L.; Brett, N.H.

A method is described for producing a dense sintered body of uranium dioxide or a mixture thereof with plutonium dioxide. Compacted uranium dioxide or a compacted uranium dioxide-plutonium dioxide mixture is heated to at least 1300 deg C in an atmosphere of carbon dioxide or carbon dioxide mixed with carbon monoxide. (R.J.S.)

Potential uranium supply from phosphoric acid: A U.S. analysis comparing solvent extraction and Ion exchange recovery

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Haeyeon; G. Eggert, Roderick; W. Carlsen, Brett

Phosphate rock contains significant amounts of uranium, although in low concentrations. Recovery of uranium as a by-product from phosphoric acid, an intermediate product produced during the recovery of phosphorus from phosphate rock, is not unprecedented. Phosphoric acid plants ceased to produce uranium as a by-product in the early 1990s with the fall of uranium prices. In the last decade, this topic has regained attention due to higher uranium prices and expected increase in demand for uranium. Our study revisits the topic and estimates how much uranium might be recoverable from current phosphoric acid production in the United States and whatmore » the associated costs might be considering two different recovery processes: solvent extraction and ion exchange. Based on U.S. phosphoric acid production in 2014, 5.5 million pounds of U 3O 8 could have been recovered, more than domestic U.S. mine production of uranium in the same year. Annualized costs for a hypothetical uranium recovery plant are US$48-66 per pound U 3O 8 for solvent extraction, the process used historically in the United States to recover uranium from phosphoric acid. For ion exchange, not yet proven at a commercial scale for uranium recovery, the estimated costs are US$33-54 per pound U 3O 8. Our results suggest that it is technically possible for the United States to recover significant quantities of uranium from current phosphoric acid production. And for this type of uranium production to be economically attractive on a large scale, either recovery costs must fall or uranium prices rise.« less

Potential uranium supply from phosphoric acid: A U.S. analysis comparing solvent extraction and Ion exchange recovery

DOE PAGES

Kim, Haeyeon; G. Eggert, Roderick; W. Carlsen, Brett; ...

2016-06-16

Phosphate rock contains significant amounts of uranium, although in low concentrations. Recovery of uranium as a by-product from phosphoric acid, an intermediate product produced during the recovery of phosphorus from phosphate rock, is not unprecedented. Phosphoric acid plants ceased to produce uranium as a by-product in the early 1990s with the fall of uranium prices. In the last decade, this topic has regained attention due to higher uranium prices and expected increase in demand for uranium. Our study revisits the topic and estimates how much uranium might be recoverable from current phosphoric acid production in the United States and whatmore » the associated costs might be considering two different recovery processes: solvent extraction and ion exchange. Based on U.S. phosphoric acid production in 2014, 5.5 million pounds of U 3O 8 could have been recovered, more than domestic U.S. mine production of uranium in the same year. Annualized costs for a hypothetical uranium recovery plant are US$48-66 per pound U 3O 8 for solvent extraction, the process used historically in the United States to recover uranium from phosphoric acid. For ion exchange, not yet proven at a commercial scale for uranium recovery, the estimated costs are US$33-54 per pound U 3O 8. Our results suggest that it is technically possible for the United States to recover significant quantities of uranium from current phosphoric acid production. And for this type of uranium production to be economically attractive on a large scale, either recovery costs must fall or uranium prices rise.« less

Extraction of uranium from tailings by sulfuric acid leaching with oxidants

NASA Astrophysics Data System (ADS)

Huang, Jing; Li, Mi; Zhang, Xiaowen; Huang, Chunmei; Wu, Xiaoyan

2017-06-01

Recovery of uranium have been performed by leaching uranium-containing tailings in sulfuric acid system with the assistance of HF, HClO4, H2O2 and MnO2. The effect of reagent dosage, sulfuric acid concentration, Liquid/solid ratio, reaction temperature and particle size on the leaching of uranium were investigated. The results show that addiction of HF, HClO4, H2O2 and MnO2 significantly increased the extraction of uranium under 1M sulphuric acid condition and under the optimum reaction conditions a dissolution fraction of 85% by HClO4, 90% by HF, 95% by H2O2 can be reached respectively. The variation of technological mineralogy properites of tailings during leaching process show that the assistants can break gangue effectively. These observations suggest that optimum oxidants could potentially influence the extraction of uranium from tailings even under dilute acid condition.

Preparation and biosorption evaluation of Bacillus subtilis/alginate–chitosan microcapsule

PubMed Central

Tong, Ke

2017-01-01

The aim of this study was to assess the effect of alginate–chitosan microcapsule on viability characteristics of Bacillus subtilis and the ability of B. subtilis/alginate–chitosan microcapsule to remove uranium ion from aqueous solution. The effects of particle size, chitosan molecular weight and inoculum density on viability characteristics were studied using alginate–chitosan microcapsule-immobilized B. subtilis experiments. In addition, the effects of pH, immobilized spherule dosage, temperature, initial uranium ion concentration and contact time on removal of uranium ion were studied using batch adsorption experiments. The results showed that alginate–chitosan microcapsule significantly improved the viability characteristics of B. subtilis and that B. subtilis/alginate–chitosan microcapsule strongly promoted uranium ion absorption. Moreover, the optimum values of pH was 6; immobilized spherule dosage was 3.5; temperature was 20°C; initial uranium ion concentration was 150 mg/L; contact time was 3 h of uranium ion absorption and the maximum adsorption capacity of uranium ion was 376.64 mg/g. PMID:28223783

40 CFR 192.10 - Applicability.

Code of Federal Regulations, 2013 CFR

2013-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Cleanup of Land and Buildings Contaminated with Residual Radioactive Materials from Inactive Uranium Processing... radioactive materials at which all or substantially all of the uranium was produced for sale to any Federal...

40 CFR 192.10 - Applicability.

Code of Federal Regulations, 2014 CFR

2014-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Cleanup of Land and Buildings Contaminated with Residual Radioactive Materials from Inactive Uranium Processing... radioactive materials at which all or substantially all of the uranium was produced for sale to any Federal...

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Size distribution of radon daughter particles in uranium mine atmospheres.

PubMed

George, A C; Hinchliffe, L; Sladowski, R

1975-06-01

The size distribution of radon daughters was measured in several uranium mines using four compact diffusion batteries and a round jet cascade impactor. Simultaneously, measurements were made of uncombined fractions of radon daughters, radon concentration, working level and particle concentration. The size distributions found for radon daughters were log normal. The activity median diameters ranged from 0.09 mum to 0.3 mum with a mean value of 0.17 mum. Geometric standard deviations were in the range from 1.3 to 4 with a mean value of 2.7. Uncombined fractions expressed in accordance with the ICRP definition ranged from 0.004 to 0.16 with a mean value of 0.04. The radon daughter sizes in these mines are greater than the sizes assumed by various authors in calculating respiratory tract dose. The disparity may reflect the widening use of diesel-powered equipment in large uranium mines.

PREPARATION OF SPHERICAL URANIUM DIOXIDE PARTICLES

DOEpatents

Levey, R.P. Jr.; Smith, A.E.

1963-04-30

This patent relates to the preparation of high-density, spherical UO/sub 2/ particles 80 to 150 microns in diameter. Sinterable UO/sub 2/ powder is wetted with 3 to 5 weight per cent water and tumbled for at least 48 hours. The resulting spherical particles are then sintered. The sintered particles are useful in dispersion-type fuel elements for nuclear reactors. (AEC)

The East Slope No. 2 uranium prospect, Piute County, Utah

USGS Publications Warehouse

Wyant, Donald Gray

1954-01-01

The secondary uranium minerals autunite, metatorbernite, uranophane(?), and schroeckingerite occur in altered hornfels at the East Slope No. 9. uranium prospect. The deposit, in sec. 6, T. 9.7 S., R. 3 W., Piute County, Utah, is about 1 mile west of the Bullion Monarch mine which is in the central producing area of the Marysvale uranium district. Hornfels, formed by contact metamorphism of rocks of the Bullion Canyon volcanics borderhug the margin of a quartz monzonite stock, is in fault contact with the later Mount Belknap rhyolite. The hornfels was intensely altered by hydrothermal solutions in pre-Mount Belknap time. Hematite-alunite-quartz-kaolinite rock, the most completely altered hornfels, is surrounded by orange to white argillized hornfels containing beidellite-montmorillonite clay, and secondary uranium minerals. The secondary uranium minerals probably have been derived from pitchblende, the primary ore mineral in other deposits of the Marysvale area. The two uranium-rich zones, 4 feet ad 5 feet thick, have been traced on the surface for 60 feet and 110 feet, respectively. Channel samples from these zones contained as much as 0.047 percent uranium. The deposit is significant because of its position outside the central producing area and because of the association of uranium minerals with alunitic rock in hydrothermally altered hornfels of volcanic rocks of early Tertiary age.

Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

NASA Astrophysics Data System (ADS)

Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

2017-01-01

The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

Isotopic analysis of uranium in natural waters by alpha spectrometry

USGS Publications Warehouse

Edwards, K.W.

1968-01-01

A method is described for the determination of U234/U238 activity ratios for uranium present in natural waters. The uranium is coprecipitated from solution with aluminum phosphate, extracted into ethyl acetate, further purified by ion exchange, and finally electroplated on a titanium disc for counting. The individual isotopes are determined by measurement of the alpha-particle energy spectrum using a high resolution low-background alpha spectrometer. Overall chemical recovery of about 90 percent and a counting efficiency of 25 percent allow analyses of water samples containing as little as 0.10 ?g/l of uranium. The accuracy of the method is limited, on most samples, primarily by counting statistics.

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; J.C. Price; R.D. Mariani

The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results andmore » conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.« less

RECOMMENDATIONS FOR UO3 PLANT BIOASSAY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbaugh, Eugene H.

2010-07-12

Alternative urine bioassay programs are described for application with decontamination and decommissioning activities at the Hanford UO3 Plant. The alternatives are based on quarterly or monthly urine bioassay for recycled uranium, assuming multiple acute inhalation intakes of recycled uranium occurring over a year. The inhalations are assumed to be 5µm AMAD particles of 80% absorption type F and 20% absorption type M. Screening levels, expressed as daily uranium mass excretion rates in urine, and the actions associated with these levels are provided for both quarterly and monthly sampling frequencies.

Occurrence and hydrogeochemistry of radiochemical constituents in groundwater of Jefferson County and surrounding areas, southwestern Montana, 2007 through 2010

USGS Publications Warehouse

Caldwell, Rodney R.; Nimick, David A.; DeVaney, Rainie M.

2014-01-01

The U.S. Geological Survey, in cooperation with Jefferson County and the Jefferson Valley Conservation District, sampled groundwater in southwestern Montana to evaluate the occurrence and concentration of naturally-occurring radioactive constituents and to identify geologic settings and environmental conditions in which elevated concentrations occur. A total of 168 samples were collected from 128 wells within Broadwater, Deer Lodge, Jefferson, Lewis and Clark, Madison, Powell, and Silver Bow Counties from 2007 through 2010. Most wells were used for domestic purposes and were primary sources of drinking water for individual households. Water-quality samples were collected from wells completed within six generalized geologic units, and analyzed for constituents including uranium, radon, gross alpha-particle activity, and gross beta-particle activity. Thirty-eight wells with elevated concentrations or activities were sampled a second time to examine variability in water quality throughout time. These water-quality samples were analyzed for an expanded list of radioactive constituents including the following: three isotopes of uranium (uranium-234, uranium-235, and uranium-238), three isotopes of radium (radium-224, radium-226, and radium-228), and polonium-210. Existing U.S. Geological Survey and Montana Bureau of Mines and Geology uranium and radon water-quality data collected as part of other investigations through 2011 from wells within the study area were compiled as part of this investigation. Water-quality data from this study were compared to data collected nationwide by the U.S. Geological Survey through 2011. Radionuclide samples for this study typically were analyzed within a few days after collection, and therefore data for this study may closely represent the concentrations and activities of water being consumed locally from domestic wells. Radioactive constituents were detected in water from every well sampled during this study regardless of location or geologic unit. Nearly 41 percent of sampled wells had at least one radioactive constituent concentration that exceeded U.S. Environmental Protection Agency drinking-water standards or screening levels. Uranium concentrations were higher than the U.S. Environmental Protection Agency maximum contaminant level (MCL) of 30 micrograms per liter in samples from 14 percent of the wells. Radon concentrations exceeded a proposed MCL of 4,000 picocuries per liter in 27 percent of the wells. Combined radium (radium-226 and radium-228) exceeded the MCL of 5 picocuries per liter in samples from 10 of 47 wells. About 40 percent (42 of 104 wells) of the wells had gross alpha-particle activities (72-hour count) at or greater than a screening level of 15 pCi/L. Gross beta-particle activity exceeded the U.S. Environmental Protection Agency 50 picocuries per liter screening level in samples from 5 of 104 wells. Maximum radium-224 and polonium-210 activities in study wells were 16.1 and 3.08 picocuries per liter, respectively; these isotopes are constituents of human-health concern, but the U.S. Environmental Protection Agency has not established MCLs for them. Radioactive constituent concentrations or activities exceeded at least one established drinking-water standard, proposed drinking-water standard, or screening level in groundwater samples from five of six generalized geologic units assessed during this study. Radioactive constituent concentrations or activities were variable not only within each geologic unit, but also among wells that were completed in the same geologic unit and in close proximity to one another. Established or proposed drinking-water standards were exceeded most frequently in water from wells completed in the generalized geologic unit that includes rocks of the Boulder batholith and other Tertiary through Cretaceous igneous intrusive rocks (commonly described as granite). Specifically, of the wells completed in the Boulder batholith and related rocks sampled as part of this study, 24 percent exceeded the MCL of 30 micrograms per liter for uranium, 50 percent exceeded the proposed alternative MCL of 4,000 picocuries per liter for radon, and 27 percent exceeded the MCL of 5 micrograms per liter for combined radium-226 and radium-228. Elevated radioactive constituent values were detected in samples representing a large range of field properties and water types. Correlations between radioactive constituents and pH, dissolved oxygen, and most major ions were not statistically significant (p-value > 0.05) or were weakly correlated with Spearman correlation coefficients (rho) ranging from -0.5 to 0.5. Moderate correlations did exist between gross beta-particle activity and potassium (rho = 0.72 to 0.82), likely because one potassium isotope (potassium-40) is a beta-particle emitter. Total dissolved solids and specific conductance also were moderately correlated (rho = 0.62 to 0.71) with gross alpha-particle and gross beta-particle activity, indicating that higher radioactivity values can be associated with higher total dissolved solids. Correlations were evaluated among radioactive constituents. Moderate to strong correlations occurred between gross alpha-particle and beta-particle activities (rho = 0.77 to 0.96) and radium isotopes (rho = 0.78 to 0.92). Correlations between gross alpha-particle activity (72-hour count) and all analyzed radioactive constituents were statistically significant (p-value Radiochemical results varied temporally in samples from several of the thirty-eight wells sampled at least twice during the study. The time between successive sampling events ranged from about 1 to 10 months for 29 wells to about 3 years for the other 9 wells. Radiochemical constituents that varied by greater than 30 percent between sampling events included uranium (29 percent of the resampled wells), and radon (11 percent of the resampled wells), gross alpha-particle activity (38 percent of the resampled wells), and gross beta-particle activity (15 percent of the resampled wells). Variability in uranium concentrations from two wells was sufficiently large that concentrations were less than the MCL in the first set of samples and greater than the MCL in the second. Sample holding times affect analytical results in this study. Gross alpha-particle and gross beta-particle activities were measured twice, 72 hours and 30 days after sample collection. Gross alpha-particle activity decreased an average of 37 percent between measurements, indicating the presence of short-lived alpha-emitting radionuclides in these samples. Gross beta-particle activity increased an average of 31 percent between measurements, indicating ingrowth of longer-lived beta-emitting radionuclides.

METHOD FOR DETERMINING THE STABILITY OF FLUOROCARBON IOLS

DOEpatents

Sheldon, Z.D.; Haendler, H.M.

1959-07-21

A method of determining the stability of a fluorocarbon oil to uranium hexafluoride is presented. The method comprises reacting a weighed sample of the oil with condensed uranium hexafluoride in a reaction zone and titrating the amount of uranium tetrafluoride produced with potassium dichromate.

Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.

1977-05-01

The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500/sup 0/C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coatedmore » in production-scale coaters, comparison of the performance of /sup 233/U-bearing particles with that of /sup 235/U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of /sup 233/U-bearing particles to be identical to that of /sup 235/U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention.« less

Concerted Uranium Research in Europe (CURE): toward a collaborative project integrating dosimetry, epidemiology and radiobiology to study the effects of occupational uranium exposure.

PubMed

Laurent, Olivier; Gomolka, Maria; Haylock, Richard; Blanchardon, Eric; Giussani, Augusto; Atkinson, Will; Baatout, Sarah; Bingham, Derek; Cardis, Elisabeth; Hall, Janet; Tomasek, Ladislav; Ancelet, Sophie; Badie, Christophe; Bethel, Gary; Bertho, Jean-Marc; Bouet, Ségolène; Bull, Richard; Challeton-de Vathaire, Cécile; Cockerill, Rupert; Davesne, Estelle; Ebrahimian, Teni; Engels, Hilde; Gillies, Michael; Grellier, James; Grison, Stephane; Gueguen, Yann; Hornhardt, Sabine; Ibanez, Chrystelle; Kabacik, Sylwia; Kotik, Lukas; Kreuzer, Michaela; Lebacq, Anne Laure; Marsh, James; Nosske, Dietmar; O'Hagan, Jackie; Pernot, Eileen; Puncher, Matthew; Rage, Estelle; Riddell, Tony; Roy, Laurence; Samson, Eric; Souidi, Maamar; Turner, Michelle C; Zhivin, Sergey; Laurier, Dominique

2016-06-01

The potential health impacts of chronic exposures to uranium, as they occur in occupational settings, are not well characterized. Most epidemiological studies have been limited by small sample sizes, and a lack of harmonization of methods used to quantify radiation doses resulting from uranium exposure. Experimental studies have shown that uranium has biological effects, but their implications for human health are not clear. New studies that would combine the strengths of large, well-designed epidemiological datasets with those of state-of-the-art biological methods would help improve the characterization of the biological and health effects of occupational uranium exposure. The aim of the European Commission concerted action CURE (Concerted Uranium Research in Europe) was to develop protocols for such a future collaborative research project, in which dosimetry, epidemiology and biology would be integrated to better characterize the effects of occupational uranium exposure. These protocols were developed from existing European cohorts of workers exposed to uranium together with expertise in epidemiology, biology and dosimetry of CURE partner institutions. The preparatory work of CURE should allow a large scale collaborative project to be launched, in order to better characterize the effects of uranium exposure and more generally of alpha particles and low doses of ionizing radiation.

Molybdenum-base cermet fuel development

NASA Astrophysics Data System (ADS)

Pilger, James P.; Gurwell, William E.; Moss, Ronald W.; White, George D.; Seifert, David A.

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. The MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermte. This cermet is to have a high matrix density (greater than or equal to 95 percent) for high strength and high thermal conductance coupled with a high particle (UN) porosity (approximately 25 percent) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous Un microspheres become available. Process development was conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and the vacuum hot press consolidation techniques. This paper summarizes the status of these activities.

Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

DOE PAGES

George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; ...

2014-12-01

Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO 2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO 2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

Uranium carbide fission target R&D for RIA - an update

NASA Astrophysics Data System (ADS)

Greene, J. P.; Levand, A.; Nolen, J.; Burtseva, T.

2004-12-01

For the Rare Isotope Accelerator (RIA) facility, ISOL targets employing refractory compounds of uranium are being developed to produce radioactive ions for post-acceleration. The availability of refractory uranium compounds in forms that have good thermal conductivity, relatively high density, and adequate release properties for short-lived isotopes remains an important issue. Investigations using commercially obtained uranium carbide material and prepared into targets involving various binder materials have been carried out at ANL. Thin sample pellets have been produced for measurements of thermal conductivity using a new method based on electron bombardment with the thermal radiation observed using a two-color optical pyrometer and performed on samples as a function of grain size, pressing pressure and sintering temperature. Manufacture of uranium carbide powder has now been achieved at ANL. Simulations have been carried out on the thermal behavior of the secondary target assembly incorporating various heat shield configurations.

Uranium isotopes quantitatively determined by modified method of atomic absorption spectrophotometry

NASA Technical Reports Server (NTRS)

Lee, G. H.

1967-01-01

Hollow-cathode discharge tubes determine the quantities of uranium isotopes in a sample by using atomic absorption spectrophotometry. Dissociation of the uranium atoms allows a large number of ground state atoms to be produced, absorbing the incident radiation that is different for the two major isotopes.

Nuclear Propulsion for Space Applications

NASA Technical Reports Server (NTRS)

Houts, M. G.; Bechtel, R. D.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

2013-01-01

Basics of Nuclear Systems: Long history of use on Apollo and space science missions. 44 RTGs and hundreds of RHUs launched by U.S. during past 4 decades. Heat produced from natural alpha (a) particle decay of Plutonium (Pu-238). Used for both thermal management and electricity production. Used terrestrially for over 65 years. Fissioning 1 kg of uranium yields as much energy as burning 2,700,000 kg of coal. One US space reactor (SNAP-10A) flown (1965). Former U.S.S.R. flew 33 space reactors. Heat produced from neutron-induced splitting of a nucleus (e.g. U-235). At steady-state, 1 of the 2 to 3 neutrons released in the reaction causes a subsequent fission in a "chain reaction" process. Heat converted to electricity, or used directly to heat a propellant. Fission is highly versatile with many applications.

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Removal of uranium (VI) from aqueous systems by nanoscale zero-valent iron particles suspended in carboxy-methyl cellulose

NASA Astrophysics Data System (ADS)

Popescu (Hoştuc), Ioana-Carmen; Filip, Petru; Humelnicu, Doina; Humelnicu, Ionel; Scott, Thomas Bligh; Crane, Richard Andrew

2013-11-01

Carboxy-methyl-cellulose (CMC), a common "delivery vehicle" for the subsurface deployment of iron nanoparticles (INP) has been tested in the current work for the removal of aqueous uranium from synthetic water samples. A comparison of the removal of aqueous uranium from solutions using carboxy-methyl-cellulose with and without iron nanoparticles (CMC-INP and CMC, respectively) was tested over a 48 h reaction period. Analysis of liquid samples using spectrophotometry determined a maximum sorption capacity of uranium, Qmax, of 185.18 mg/g and 322.58 mg/g for CMC and CMC-INP respectively, providing strong evidence of an independent aqueous uranium removal ability exhibited by CMC. The results point out that CMC provides an additional capacity for aqueous uranium removal. Further tests are required to determine whether similar behaviour will be observed for other aqueous contaminant species and if the presence of CMC within a INP slurry inhibits or aids the reactivity, reductive capacity and affinity of INP for aqueous contaminant removal.

PROCESS OF PREPARING A FLUORIDE OF TETRAVLENT URANIUM

DOEpatents

Wheelwright, E.J.

1959-02-17

A method is described for producing a fluoride salt pf tetravalent uranium suitable for bomb reduction to metallic uranium. An aqueous solution of uranyl nitrate is treated with acetic acid and a nitrite-suppressor and then contacted with metallic lead whereby uranium is reduced from the hexavalent to the tetravalent state and soluble lead acetate is formed. Sulfate ions are then added to the solution to precipitate and remove the lead values. Hydrofluoric acid and alkali metal ions are then added causing the formation of an alkali metal uranium double-fluoride in which the uranium is in the tetravalent state. After recovery, this precipitate is suitable for using in the limited production of metallic uranium.

Nuclear reactor target assemblies, nuclear reactor configurations, and methods for producing isotopes, modifying materials within target material, and/or characterizing material within a target material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, James J.; Wall, Donald; Wittman, Richard S.

Target assemblies are provided that can include a uranium-comprising annulus. The assemblies can include target material consisting essentially of non-uranium material within the volume of the annulus. Reactors are disclosed that can include one or more discrete zones configured to receive target material. At least one uranium-comprising annulus can be within one or more of the zones. Methods for producing isotopes within target material are also disclosed, with the methods including providing neutrons to target material within a uranium-comprising annulus. Methods for modifying materials within target material are disclosed as well as are methods for characterizing material within a targetmore » material.« less

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

Groundwater quality of the Gulf Coast aquifer system, Houston, Texas, 2010

USGS Publications Warehouse

Oden, Jeannette H.; Brown, Dexter W.; Oden, Timothy D.

2011-01-01

Gross alpha-particle activities and beta-particle activities for all 47 samples were analyzed at 72 hours after sample collection and again at 30 days after sample collection, allowing for the measurement of the activity of short-lived isotopes. Gross alpha-particle activities reported in this report were not adjusted for activity contributions by radon or uranium and, therefore, are conservatively high estimates if compared to the U.S. Environmental Protection Agency National Primary Drinking Water Regulation for adjusted gross alpha-particle activity. The gross alpha-particle activities at 30 days in the samples ranged from R0.60 to 25.5 picocuries per liter and at 72 hours ranged from 2.58 to 39.7 picocuries per liter, and the "R" preceding the value of 0.60 picocuries per liter refers to a nondetected result less than the sample-specific critical level. Gross beta-particle activities measured at 30 days ranged from 1.17 to 14.4 picocuries per liter and at 72 hours ranged from 1.97 to 4.4 picocuries per liter. Filtered uranium was detected in quantifiable amounts in all of the 47 wells sampled. The uranium concentrations ranged from 0.03 to 42.7 micrograms per liter. One sample was analyzed for carbon-14, and the amount of modern atmospheric carbon was reported as 0.2 percent. Six source-water samples collected from municipal supply wells were analyzed for radium-226, and all of the concentrations were considered detectable concentrations (greater than their associated sample-specific critical level). Three source-water samples collected were analyzed for radon-222, and all of the concentrations were substantially greater than the associated sample-specific critical level.

Raoultella sp. SM1, a novel iron-reducing and uranium-precipitating strain.

PubMed

Sklodowska, Aleksandra; Mielnicki, Sebastian; Drewniak, Lukasz

2018-03-01

The main aim of this study was the characterisation of novel Raoutella isolate, an iron-reducing and uranium-precipitating strain, originating from microbial mats occurring in the sediments of a closed down uranium mine in Kowary (SW Poland). Characterisation was done in the context of its potential role in the functioning of these mats and the possibility to use them in uranium removal/recovery processes. In our experiment, we observed the biological precipitation of iron and uranium's secondary minerals containing oxygen, potassium, sodium and phosphor, which were identified as ningyoite-like minerals. The isolated strain, Raoultella sp. SM1, was also able to dissimilatory reduce iron (III) and uranium (VI) in the presence of citrate as an electron donor. Our studies allowed us to characterise a new strain which may be used as a model microorganism in the study of Fe and U respiratory processes and which may be useful in the bioremediation of uranium-contaminated waters and sediments. During this process, uranium may be immobilised in ningyoite-like minerals and can then be recovered in nano/micro-particle form, which may be easily transformed to uraninite. Copyright © 2017 Elsevier Ltd. All rights reserved.

Depleted uranium instead of lead in munitions: the lesser evil.

PubMed

Jargin, Sergei V

2014-03-01

Uranium has many similarities to lead in its exposure mechanisms, metabolism and target organs. However, lead is more toxic, which is reflected in the threshold limit values. The main potential hazard associated with depleted uranium is inhalation of the aerosols created when a projectile hits an armoured target. A person can be exposed to lead in similar ways. Accidental dangerous exposures can result from contact with both substances. Encountering uranium fragments is of minor significance because of the low penetration depth of alpha particles emitted by uranium: they are unable to penetrate even the superficial keratin layer of human skin. An additional cancer risk attributable to the uranium exposure might be significant only in case of prolonged contact of the contaminant with susceptible tissues. Lead intoxication can be observed in the wounded, in workers manufacturing munitions etc; moreover, lead has been documented to have a negative impact on the intellectual function of children at very low blood concentrations. It is concluded on the basis of the literature overview that replacement of lead by depleted uranium in munitions would be environmentally beneficial or largely insignificant because both lead and uranium are present in the environment.

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

NASA Astrophysics Data System (ADS)

Towell, Rusty; Niffte Collaboration

2015-10-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

Uranium ores and depleted uranium in the environment, with a reference to uranium in the biosphere from the Erzgebirge/Sachsen, Germany.

PubMed

Meinrath, A; Schneider, P; Meinrath, G

2003-01-01

The Erzgebirge ('Ore Mountains') area in the eastern part of Germany was a major source of uranium for Soviet nuclear programs between 1945 and 1989. During this time, the former German Democratic Republic became the third largest uranium producer in the world. The high abundance of uranium in the geological formations of the Erzgebirge are mirrored in the discovery of uranium by M. Klaproth close to Freiberg City in 1789 and the description of the so-called 'Schneeberg' disease, lung cancer caused in miners by the accumulation of the uranium decay product, radon, in the subsurfaces of shafts. Since 1991, remediation and mitigation of uranium at production facilities, rock piles and mill tailings has taken place. In parallel, efforts were initiated to assess the likely adverse effects of uranium mining to humans. The costs of these activities amount to about 6.5 10(9) Euro. A comparison with concentrations of depleted uranium at certain sites is given.

Process for producing an aggregate suitable for inclusion into a radiation shielding product

DOEpatents

Lessing, Paul A.; Kong, Peter C.

2000-01-01

The present invention is directed to methods for converting depleted uranium hexafluoride to a stable depleted uranium silicide in a one-step reaction. Uranium silicide provides a stable aggregate material that can be added to concrete to increase the density of the concrete and, consequently, shield gamma radiation. As used herein, the term "uranium silicide" is defined as a compound generically having the formula U.sub.x Si.sub.y, wherein the x represents the molecules of uranium and the y represent the molecules of silicon. In accordance with the present invention, uranium hexafluoride is converted to a uranium silicide by contacting the uranium hexafluoride with a silicon-containing material at a temperature in a range between about 1450.degree. C. and about 1750.degree. C. The stable depleted uranium silicide is included as an aggregate in a radiation shielding product, such as a concrete product.

Chemical analysis of water samples and geophysical logs from cored test holes drilled in the central Oklahoma Aquifer, Oklahoma

USGS Publications Warehouse

Schlottmann, Jamie L.; Funkhouser, Ron A.

1991-01-01

Chemical analyses of water from eight test holes and geophysical logs for nine test holes drilled in the Central Oklahoma aquifer are presented. The test holes were drilled to investigate local occurrences of potentially toxic, naturally occurring trace substances in ground water. These trace substances include arsenic, chromium, selenium, residual alpha-particle activities, and uranium. Eight of the nine test holes were drilled near wells known to contain large concentrations of one or more of the naturally occurring trace substances. One test hole was drilled in an area known to have only small concentrations of any of the naturally occurring trace substances.Water samples were collected from one to eight individual sandstone layers within each test hole. A total of 28 water samples, including four duplicate samples, were collected. The temperature, pH, specific conductance, alkalinity, and dissolved-oxygen concentrations were measured at the sample site. Laboratory determinations included major ions, nutrients, dissolved organic carbon, and trace elements (aluminum, arsenic, barium, beryllium, boron, cadmium, chromium, hexavalent chromium, cobalt, copper, iron, lead, lithium, manganese, mercury, molybdenum, nickel, selenium, silver, strontium, vanadium and zinc). Radionuclide activities and stable isotope (5 values also were determined, including: gross-alpha-particle activity, gross-beta-particle activity, radium-226, radium-228, radon-222, uranium-234, uranium-235, uranium-238, total uranium, carbon-13/carbon-12, deuterium/hydrogen-1, oxygen-18/oxygen-16, and sulfur-34/sulfur-32. Additional analyses of arsenic and selenium species are presented for selected samples as well as analyses of density and iodine for two samples, tritium for three samples, and carbon-14 for one sample.Geophysical logs for most test holes include caliper, neutron, gamma-gamma, natural-gamma logs, spontaneous potential, long- and short-normal resistivity, and single-point resistance. Logs for test-hole NOTS 7 do not include long- and short-normal resistivity, spontaneous-potential, or single-point resistivity. Logs for test-hole NOTS 7A include only caliper and natural-gamma logs.

APPARATUS FOR CHARGING A RECEPTACLE WITH A DENSE SUBLIMATE FORM OF URANIUM CHLORIDE

DOEpatents

Davidson, P.H.

1959-08-18

An apparatus for filling a tubular storage receptacle with a dense massive form of uranium chloride is described. The apparatus includes an evacuated housing divided into a vaporizing chamber and a portion adapted to receive the receptacle. A nozzle conducts vaporized uranium chloride from the chamber to the interior of the receptacle. The nozzle is withdrawable to progressively deposit the uranium chloride under controlled conditions to produce a dense sublimate which fills the receptacle.

Electronic structure and fine structural features of the air-grown UNxOy on nitrogen-rich uranium nitride

NASA Astrophysics Data System (ADS)

Long, Zhong; Zeng, Rongguang; Hu, Yin; Liu, Jing; Wang, Wenyuan; Zhao, Yawen; Luo, Zhipeng; Bai, Bin; Wang, Xiaofang; Liu, Kezhao

2018-06-01

Oxide formation on surface of nitrogen-rich uranium nitride film/particles was investigated using X-ray photoelectron spectroscopy (XPS), auger electron spectroscopy (AES), aberration-corrected transmission electron microscopy (TEM), and high-angle annular dark-field scanning transmission electron microscopy (HAADF-STEM) coupled with electron energy-loss spectroscopy (EELS). XPS and AES studies indicated that the oxidized layer on UN2-x film is ternary compound uranium oxynitride (UNxOy) in 5-10 nm thickness. TEM/HAADF-STEM and EELS studies revealed the UNxOy crystallizes in the FCC CaF2-type structure with the lattice parameter close to the CaF2-type UN2-x matrix. The work can provide further information to the oxidation mechanism of uranium nitride.

Assessing the deposition of radon progeny from a uranium glass necklace.

PubMed

Hansen, M F; Moss, G R

2015-06-01

Could jewellery made from uranium glass beads pose an increased risk to skin cancer? The literature Eatough (Alpha-particle dosimetry for the basal layer of the skin and the radon progeny (218)Po and (214)Po. Phys. Med. Biol. 1997; 42: 1899-1911.) suggests that the alphas from the short-lived radon daughters, (218)Po and (214)Po, may reach the basal layer of the epidermis, which is believed to be important in the induction of skin cancers. The deposition of the alphas from the (218)Po and (214)Po daughters was investigated using PADC detector material. The expectation would be that no alpha particles would penetrate through the dead skin layer, assuming the average of 70 microns used in radiation protection, but the skin around the collar bone could potentially be thinner than the assumed average. It should be noticed that by inserting a slice of pig skin in between the necklace and the PADC, no great excess of alpha tracks were seen after 1 week of exposure in the freezer. There was, however, a clear signal through the pig skin from beta particles, confirming the potential of a uranium bead necklace posing a health risk. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

75 FR 76050 - Notice of the Nuclear Regulatory Commission Consent to Indirect Change of Control and Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

... Ranch in situ recovery (ISR) project in Johnson and Campbell Counties, Wyoming. The project is currently in operating status, but is not producing uranium at this time. Materials License SUA-1569 authorizes Uranium One Americas, Inc., to possess uranium and byproduct material at its Moore Ranch ISR Project in...

RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

DOEpatents

Fromm, L.W. Jr.

1959-09-01

An electrolytic process for recovering uranium from a graphite fuel element is described. The uraniumcontaining graphite body is disposed as the anode of a cell containing a nitric acid electrolyte and a 5 amp/cm/sup 2/ current passed to induce a progressive disintegration of the graphite body. The dissolved uranium is quickly and easily separated from the resulting graphite particles by simple mechanical means, such as centrifugation, filtration, and decontamination.

The manufacture of LEU fuel elements at Dounreay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gibson, J.

1997-08-01

Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

Remediation of soils contaminated with particulate depleted uranium by multi stage chemical extraction.

PubMed

Crean, Daniel E; Livens, Francis R; Sajih, Mustafa; Stennett, Martin C; Grolimund, Daniel; Borca, Camelia N; Hyatt, Neil C

2013-12-15

Contamination of soils with depleted uranium (DU) from munitions firing occurs in conflict zones and at test firing sites. This study reports the development of a chemical extraction methodology for remediation of soils contaminated with particulate DU. Uranium phases in soils from two sites at a UK firing range, MOD Eskmeals, were characterised by electron microscopy and sequential extraction. Uranium rich particles with characteristic spherical morphologies were observed in soils, consistent with other instances of DU munitions contamination. Batch extraction efficiencies for aqueous ammonium bicarbonate (42-50% total DU extracted), citric acid (30-42% total DU) and sulphuric acid (13-19% total DU) were evaluated. Characterisation of residues from bicarbonate-treated soils by synchrotron microfocus X-ray diffraction and X-ray absorption spectroscopy revealed partially leached U(IV)-oxide particles and some secondary uranyl-carbonate phases. Based on these data, a multi-stage extraction scheme was developed utilising leaching in ammonium bicarbonate followed by citric acid to dissolve secondary carbonate species. Site specific U extraction was improved to 68-87% total U by the application of this methodology, potentially providing a route to efficient DU decontamination using low cost, environmentally compatible reagents. Copyright © 2013 The Authors. Published by Elsevier B.V. All rights reserved.

Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317more » containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).« less

Analysis of a Uranium Oxide Sample Interdicted in Slovakia (FSC 12-3-1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borg, Lars E.; Dai, Zurong; Eppich, Gary R.

2014-01-17

We provide a concise summary of analyses of a natural uranium sample seized in Slovakia in November 2007. Results are presented for compound identification, water content, U assay, trace element abundances, trace organic compounds, isotope compositions for U, Pb, Sr and O, and age determination using the 234U – 230Th and 235U – 231Pa chronometers. The sample is a mixture of two common uranium compounds - schoepite and uraninite. The uranium isotope composition is indistinguishable from natural; 236U was not detected. The O, Sr and Pb isotope compositions and trace element abundances are unremarkable. The 234U – 230Th chronometer givesmore » an age of 15.5 years relative to the date of analysis, indicating the sample was produced in January 1997. A comparison of the data for this sample with data in the Uranium Sourcing database failed to find a match, indicating the sample was not produced at a facility represented in the database.« less

A top-down assessment of energy, water and land use in uranium mining, milling, and refining

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Land, water and energy use are key measures of the sustainability of uranium production into the future. As the most attractive, accessible deposits are mined out, future discoveries may prove to be significantly, perhaps unsustainably, more intensive consumers of environmental resources. A number of previous attempts have been made to provide empirical relationships connecting these environmental impact metrics to process variables such as stripping ratio and ore grade. These earlier attempts were often constrained by a lack of real world data and perform poorly when compared against data from modern operations. This paper conditions new empirical models of energy, watermore » and land use in uranium mining, milling, and refining on contemporary data reported by operating mines. It shows that, at present, direct energy use from uranium production represents less than 1% of the electrical energy produced by the once-through fuel cycle. Projections of future energy intensity from uranium production are also possible by coupling the empirical models with estimates of uranium crustal abundance, characteristics of new discoveries, and demand. The projections show that even for the most pessimistic of scenarios considered, by 2100, the direct energy use from uranium production represents less than 3% of the electrical energy produced by the contemporary once-through fuel cycle.« less

Geological and geochemical aspects of uranium deposits. A selected, annotated bibliography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garland, P.A.; Thomas, J.M.; Brock, M.L.

1980-06-01

A bibliography of 479 references encompassing the fields of uranium and thorium geochemistry and mineralogy, geology of uranium deposits, uranium mining, and uranium exploration techniques has been compiled by the Ecological Sciences Information Center of Oak Ridge National Laboratory. The bibliography was produced for the National Uranium Resource Evaluation Program, which is funded by the Grand Junction Office of the Department of Energy. The references contained in the bibliography have been divided into the following eight subject categories: (1) geology of deposits, (2) geochemistry, (3) genesis O deposits, (4) exploration, (5) mineralogy, (6) uranium industry, (7) reserves and resources, andmore » (8) geology of potential uranium-bearing areas. All categories specifically refer to uranium and thorium; the last category contains basic geologic information concerning areas which the Grand Junction Office feels are particularly favorable for uranium deposition. The references are indexed by author, geographic location, quadrangle name, geoformational feature, taxonomic name, and keyword.« less

Bioreduction of U(VI)-Phthalate to a Polymeric U(IV)-Phthalate Colloid

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vazquez, G.; Dodge, C; Francis, A

2009-01-01

Phthalic acid, a ubiquitous organic ligand, formed soluble mono- and biligand complexes with a uranyl ion that was then reduced to a U(IV)-phthalate by a Clostridium species under anaerobic conditions. We confirmed the reduction of the hexavalent uranium to the tetravalent oxidation state by UV-vis absorption and X-ray absorption near edge structure spectroscopy. Sequential micro- and ultrafiltration of the solution revealed that the bioreduced uranium was present as a colloid with particles between 0.03 and 0.45 {mu}m. Analysis with extended X-ray absorption fine structure revealed the association of the reduced uranium with the phthalic acid as a repeating biligand 1:2more » U(IV):phthalic acid polymer. This is the first report of the formation of a U(IV) complexed to two phthalic acid molecules in the form of a polymeric colloid. Although it was proposed that the bioreduction and the precipitation of uranium might be an invaluable strategy to immobilize uranium in contaminated environments, our results suggest that the organic ligands present there might hinder the precipitation of the bioreduced uranium under anaerobic conditions and, thereby, enhance its environmental mobility as uranium organic complexes or colloids.« less

Localized corrosion of 316L stainless steel in tritiated water containing aggressive radiolytic and decomposition products at different temperatures

NASA Astrophysics Data System (ADS)

Bellanger, G.

2008-02-01

Tritium is one of the more important radionuclides used in nuclear industry as plutonium and uranium. The tritium in tritiated water always causes difficulties in nuclear installations, including equipment corrosion. Moreover, with tritiated water there are, in addition, the radiolytic and decomposition products such as hydrogen peroxide formed during decay, chloride ions produced by degradation of organic seals and oils used for tightness and pumping, and acid pH produced by excitation of nitrogen in air by the β - particle. Highly concentrated tritiated water releases energy and its temperature is about 80 °C, moreover heating is necessary in the tritium processes. These conditions highly facilitate the corrosion of stainless steels by pitting and crevice attack. Corrosion tests were performed by electrochemical analysis methods and by visual inspection of the surface of stainless steel.

Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu

2011-03-30

Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO{sub 4}) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1x10{sup 13} ncm{sup -2}s{sup -1}. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emissionmore » of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.« less

Physico-Chemical Heterogeneity of Organic-Rich Sediments in the Rifle Aquifer, CO: Impact on Uranium Biogeochemistry.

PubMed

Janot, Noémie; Lezama Pacheco, Juan S; Pham, Don Q; O'Brien, Timothy M; Hausladen, Debra; Noël, Vincent; Lallier, Florent; Maher, Kate; Fendorf, Scott; Williams, Kenneth H; Long, Philip E; Bargar, John R

2016-01-05

The Rifle alluvial aquifer along the Colorado River in west central Colorado contains fine-grained, diffusion-limited sediment lenses that are substantially enriched in organic carbon and sulfides, as well as uranium, from previous milling operations. These naturally reduced zones (NRZs) coincide spatially with a persistent uranium groundwater plume. There is concern that uranium release from NRZs is contributing to plume persistence or will do so in the future. To better define the physical extent, heterogeneity and biogeochemistry of these NRZs, we investigated sediment cores from five neighboring wells. The main NRZ body exhibited uranium concentrations up to 100 mg/kg U as U(IV) and contains ca. 286 g of U in total. Uranium accumulated only in areas where organic carbon and reduced sulfur (as iron sulfides) were present, emphasizing the importance of sulfate-reducing conditions to uranium retention and the essential role of organic matter. NRZs further exhibited centimeter-scale variations in both redox status and particle size. Mackinawite, greigite, pyrite and sulfate coexist in the sediments, indicating that dynamic redox cycling occurs within NRZs and that their internal portions can be seasonally oxidized. We show that oxidative U(VI) release to the aquifer has the potential to sustain a groundwater contaminant plume for centuries. NRZs, known to exist in other uranium-contaminated aquifers, may be regionally important to uranium persistence.

URANIUM RECOVERY AND PURIFICATION PROCESS AND PRODUCTION OF HIGH PURITY URANIUM TETRAFLUORIDE

DOEpatents

Bailes, R.H.; Long, R.S.; Grinstead, R.R.

1957-09-17

A process is described wherein an anionic exchange technique is employed to separate uramium from a large variety of impurities. Very efficient and economical purification of contamimated uranium can be achieved by treatment of the contaminated uranium to produce a solution containing a high concentration of chloride. Under these conditions the uranium exists as an aniomic chloride complex. Then the uranium chloride complex is adsorbed from the solution on an aniomic exchange resin, whereby a portion of the impurities remain in the solution and others are retained with the uramium by the resin. The adsorbed impurities are then removed by washing the resin with pure concentrated hydrochloric acid, after which operation the uranium is eluted with pure water yielding an acidic uranyl chloride solution of high purity.

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

DOEpatents

Piper, R.D.

1962-09-01

A process is given for making uranium metal from oxidic material by electrolytic deposition on the cathode. The oxidic material admixed with two moles of carbon per one mole of uranium dioxide forms the anode, and the electrolyte is a mixture of from 40 to 75% of calcium fluoride or barium fluoride, 15 to 45% of uranium tetrafluoride, and from 10 to 20% of lithium fluoride or magnesium fluoride; the temperature of the electrolyte is between 1150 and 1175 deg C. (AEC)

FABRICATION OF URANIUM-ALUMINUM ALLOYS

DOEpatents

Saller, H.A.

1959-12-15

A process is presented for producing a workable article of a uranium- aluminum alloy in which the uranium content is between 14 and 70% by weight; aluminum powder and powdered UAl/sub 2/, UAl/sub 3/, UAl/sub 5/, or UBe/sub 9/ are mixed, and the mixture is compressed into the shape desired and sintered at between 450 and 600 deg C.

A model of early formation of uranium molecular oxides in laser-ablated plasmas

NASA Astrophysics Data System (ADS)

Finko, Mikhail; Curreli, Davide; Azer, Magdi; Weisz, David; Crowhurst, Jonathan; Rose, Timothy; Koroglu, Batikan; Radousky, Harry; Zaug, Joseph; Armstrong, Mike

2017-10-01

An important problem within the field of nuclear forensics is fractionation: the formation of post-detonation nuclear debris whose composition does not reflect that of the source weapon. We are investigating uranium fractionation in rapidly cooling plasma using a combined experimental and modeling approach. In particular, we use laser ablation of uranium metal samples to produce a low-temperature plasma with physical conditions similar to a condensing nuclear fireball. Here we present a first plasma-chemistry model of uranium molecular species formation during the early stage of laser ablated plasma evolution in atmospheric oxygen. The system is simulated using a global kinetic model with rate coefficients calculated according to literature data and the application of reaction rate theory. The model allows for a detailed analysis of the evolution of key uranium molecular species and represents the first step in producing a uranium fireball model that is kinetically validated against spatially and temporally resolved spectroscopy measurements. This project was sponsored by the DoD, Defense Threat Reduction Agency, Grant HDTRA1-16- 1-0020. This work was performed in part under the auspices of the U.S. DoE by Lawrence Livermore National Laboratory under Contract DE-AC52- 07NA27344.

BLENDING LOW ENRICHED URANIUM WITH DEPLETED URANIUM TO CREATE A SOURCE MATERIAL ORE THAT CAN BE PROCESSED FOR THE RECOVERY OF YELLOWCAKE AT A CONVENTIONAL URANIUM MILL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schutt, Stephen M.; Hochstein, Ron F.; Frydenlund, David C.

2003-02-27

Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials atmore » a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.« less

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

Studies on supercritical hydrothermal syntheses of uranium and lanthanide oxide particles and their reaction mechanisms

NASA Astrophysics Data System (ADS)

Hwang, DongKi; Tsukahara, Takehiko; Tanaka, Kosuke; Osaka, Masahiko; Ikeda, Yasuhisa

2015-11-01

In order to develop preparation method of raw metal oxide particles for low decontaminated MOX fuels by supercritical hydrothermal (SH) treatments, we have investigated behavior of aqueous solutions dissolving U(VI), Ln(III) (Ln: lanthanide = Ce, Pr, Nd, Sm, Tb), Cs(I), and Sr(II) nitrate or chloride compounds under SH conditions (temperature = 400-500 °C, pressure = 30-40 MPa). As a result, it was found that Ln(NO3)3 (Ln = Ce, Pr, Tb) compounds produce LnO2, that Ln(NO3)3 (Ln = Nd, Sm) compounds are hardly converted to their oxides, and that LnCl3 (Ln = Ce, Pr, Nd, Sm, Tb), CsNO3, and Sr(NO3)2 do not form their oxide compounds. Furthermore, HNO2 species were detected in the liquid phase obtained after treating HNO3 aqueous solutions containing Ln(NO3)3 (Ln = Ce, Pr, Tb) under SH conditions, and also NO2 and NO compounds were found to be produced by decomposition of HNO3. From these results, it was proposed that the Ln oxide (LnO2) particles are directly formed with oxidation of Ln(III) to Ln(IV) by HNO3 and HNO2 species in the SH systems. Moreover, the uranyl ions were found to form U3O8 and UO3 depending on the concentration of HNO3. From these results, it is expected that the raw metal oxide particles for low decontaminated MOX fuels are efficiently prepared by the SH method.

Uranium from German Nuclear Power Projects of the 1940s— A Nuclear Forensic Investigation

PubMed Central

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-01-01

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel.3b,d, 4 Through measurement of the 230Th/234U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the 87Sr/86Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of 236U and 239Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. PMID:26501922

FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

DOEpatents

Creutz, E.C.

1959-01-27

A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

Chemistry and particle track studies of Apollo 14 glasses.

NASA Technical Reports Server (NTRS)

Glass, B. P.; Storzer, D.; Wagner, G. A.

1972-01-01

The abundance and the composition of Apollo 14 glasses have been studied. Glass particles were analyzed for Si, Ti, Al, Fe, Mn, Mg, Na, and K by electron microprobe analysis. The refractive indices of 26 particles were determined by the oil immersion method. Track analyses have been carried out in order to determine the uranium content and the radiation history of glass particles. The proper identification of galactic and solar flare nuclei tracks makes it possible to estimated residence times of the glass particles in the top layer of the lunar soil.

Kinetic, equilibrium and thermodynamic studies on sorption of uranium and thorium from aqueous solutions by a selective impregnated resin containing carminic acid.

PubMed

Rahmani-Sani, Abolfazl; Hosseini-Bandegharaei, Ahmad; Hosseini, Seyyed-Hossein; Kharghani, Keivan; Zarei, Hossein; Rastegar, Ayoob

2015-04-09

In this work, the removal of uranium and thorium ions from aqueous solutions was studied by solid-liquid extraction using an advantageous extractant-impregnated resin (EIR) prepared by loading carminic acid (CA) onto Amberlite XAD-16 resin beads. Batch sorption experiments using CA/XAD-16 beads for the removal of U(VI) and Th(IV) ions were carried out as a function of several parameters, like equilibration time, metal ion concentration, etc. The equilibrium data obtained from the sorption experiments were adjusted to the Langmuir isotherm model and the calculated maximum sorption capacities in terms of monolayer sorption were in agreement with those obtained from the experiments. The experimental data on the sorption behavior of both metal ions onto the EIR beads fitted well in both Bangham and intra-particle diffusion kinetic models, indicating that the intra-particle diffusion is the rate-controlling step. The thermodynamic studies at different temperatures revealed the feasibility and the spontaneous nature of the sorption process for both uranium and thorium ions. Copyright © 2014 Elsevier B.V. All rights reserved.

METHOD OF DEHYDRATING URANIUM TETRAFLUORIDE

DOEpatents

Davis, J.O.; Fogel, C.C.; Palmer, W.E.

1962-12-18

Drying and dehydration of aqueous-precipitated uranium tetrafluoride are described. The UF/sub 4/ which normally contains 3 to 4% water, is dispersed into the reaction zone of an operating reactor wherein uranium hexafluoride is being reduced to UF/sub 4/ with hydrogen. The water-containing UF/sub 4/ is dried and blended with the UF/sub 4/ produced in the reactor without interfering with the reduction reaction. (AEC)

Nuclear Fuel Reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Visible spectral power emitted from a laser produced uranium plasma

NASA Technical Reports Server (NTRS)

Williams, M. D.; Jalufka, N. W.

1975-01-01

The development of plasma-core nuclear reactors for advanced terrestrial and space-power sources is researched. Experimental measurements of the intensity and the spectral distribution of radiation from a nonfissioning uranium plasma are reported.

40 CFR 440.30 - Applicability; description of the uranium, radium and vanadium ores subcategory.

Code of Federal Regulations, 2014 CFR

2014-07-01

... vanadium ores are produced; and (b) mills using the acid leach, alkaline leach, or combined acid and alkaline leach process for the extraction of uranium, radium and vanadium. Only vanadium byproduct...

40 CFR 440.30 - Applicability; description of the uranium, radium and vanadium ores subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... vanadium ores are produced; and (b) mills using the acid leach, alkaline leach, or combined acid and alkaline leach process for the extraction of uranium, radium and vanadium. Only vanadium byproduct...

40 CFR 440.30 - Applicability; description of the uranium, radium and vanadium ores subcategory.

Code of Federal Regulations, 2013 CFR

2013-07-01

... vanadium ores are produced; and (b) mills using the acid leach, alkaline leach, or combined acid and alkaline leach process for the extraction of uranium, radium and vanadium. Only vanadium byproduct...

Uranium plasma emission at gas-core reaction conditions

NASA Technical Reports Server (NTRS)

Williams, M. D.; Jalufka, N. W.; Hohl, F.; Lee, J. H.

1976-01-01

The results of uranium plasma emission produced by two methods are reported. For the first method a ruby laser was focused on the surface of a pure U-238 sample to create a plasma plume with a peak plasma density of about 10 to the 20th power/cu cm and a temperature of about 38,600 K. The absolute intensity of the emitted radiation, covering the range from 300 to 7000 A was measured. For the second method, the uranium plasma was produced in a 20 kilovolt, 25 kilojoule plasma-focus device. The 2.5 MeV neutrons from the D-D reaction in the plasma focus are moderated by polyethylene and induce fissions in the U-235. Spectra of both uranium plasmas were obtained over the range from 30 to 9000 A. Because of the low fission yield the energy input due to fissions is very small compared to the total energy in the plasma.

Electrolytic process for preparing uranium metal

DOEpatents

Haas, Paul A.

1990-01-01

An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

Antiproton Production in Relativistic Heavy Ion Collisions

NASA Astrophysics Data System (ADS)

Greene, Senta Victoria

The E814 collaboration has made a systematic study of antiproton production in collisions of ^ {28}Si ions at 14.6 GeV per nucleon with targets of Pb, Cu, and Al. This study was motivated by the expectation that antiprotons will be a useful probe of the system produced in relativistic heavy ion collisions. The large annihilation cross section for antiprotons makes the antiproton survival probability sensitive to the baryon density of the system in which they are created. It has also been suggested that a transition to the quark-gluon plasma phase may produce an enhancement of antibaryon production. The E814 spectrometer consists of three tracking chambers for momentum measurement, a scintillator hodoscope to measure charge and time of flight, and a sampling calorimeter. The spectrometer accepts all particles produced within a rectangular aperture centered on the beam axis, with delta theta_{x}=37.6mr and deltatheta_{y}=24.1mr. A trigger based on the flight time of particles through the spectrometer enhances the selection of events which produce negatively charged particles having a rapidity within 0.5 units of the center of mass rapidity. Measurements of the antiproton yield per interaction and the invariant cross section for production at zero degrees are presented and discussed. The time-of-flight trigger allows for an unbiased measurement of the probability to produce antiprotons as a function of the impact parameter of the collision. Several measures of collision centrality are used. The energy produced transverse to the beam direction is measured with the target calorimeter, an array of NaI crystals surrounding the target assembly with a pseudorapidity coverage of -0.5

Simplified behaviors from increased heterogeneity: II. 3-D uranium transport at the decimeter scale and intertank comparisons.

PubMed

Miller, Andrew W; Rodriguez, Derrick R; Honeyman, Bruce D

2013-05-01

Upscaling from bench scale systems to field scale systems incorporates physical and chemical heterogeneities from atomistic up to field scales. Heterogeneities of intermediate scale (~10(-1) m) are impossible to incorporate in a bench scale experiment. To transcend these scale discrepancies, this second in a pair of papers presents results from an intermediate scale, 3-D tank experiment completed using five different particle sizes of uranium contaminated sediment from a former uranium mill field site. The external dimensions of the tank were 2.44 m×0.61 m×0.61 m (L×H×W). The five particle sizes were packed in a heterogeneous manner using roughly 11 cm cubes. Small groundwater wells were installed for spatial characterization of chemical gradients and flow parameters. An approximately six month long bromide tracer test was used for flow field characterization. Within the flow domain, local uranium breakthrough curves exhibited a wide range of behaviors. However, the global effluent breakthrough curve was smooth, and not unlike breakthrough curves observed in column scale experiments. This paper concludes with an inter-tank comparison of all three experimental systems presented in this pair of papers. Although there is a wide range of chemical and physical variability between the three tanks, major chemical constituent behaviors are often quite similar or even identical. Copyright © 2013 Elsevier B.V. All rights reserved.

Spent nuclear fuel recycling with plasma reduction and etching

DOEpatents

Kim, Yong Ho

2012-06-05

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF₆ and PuF₄. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF₆ (gas) that is formed to decompose back to PuF₄ (solid), and b) to maintain stability of the UF₆. Uranium (in the form of gaseous UF₆) is easily extracted and separated from the plutonium (in the form of solid PuF₄). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF₆ (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF₄ or NF₃ as the fluorine sources instead of F₂ or HF.

Measuring aerosols generated inside armoured vehicles perforated by depleted uranium ammunition.

PubMed

Parkhurst, M A

2003-01-01

In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the US Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects on soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterised. A robust sampling system was designed to collect aerosols in this difficult environment and monitor continuously the sampler flow rates. The aerosol samplers selected for these tests included filter cassettes, cascade impactors, a five-stage cyclone and a moving filter. Sampler redundancy was an integral part of the sampling system to offset losses from fragment damage. Wipe surveys and deposition trays collected removable deposited particulate matter. Interior aerosols were analysed for uranium concentration and particle size distribution as a function of time. They were also analysed for uranium oxide phases, particle morphology and dissolution in vitro. These data, currently under independent peer review, will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.

Subcellular distribution of uranium in the roots of Spirodela punctata and surface interactions

NASA Astrophysics Data System (ADS)

Nie, Xiaoqin; Dong, Faqin; Liu, Ning; Liu, Mingxue; Zhang, Dong; Kang, Wu; Sun, Shiyong; Zhang, Wei; Yang, Jie

2015-08-01

The subcellular distribution of uranium in roots of Spirodela punctata (duckweed) and the process of surface interaction were studied upon exposure to U (0, 5-200 mg/L) at pH 5. The concentration of uranium in each subcelluar fraction increased significantly with increasing solution U level, after 200 mg/L uranium solution treatment 120 h, the proportion of uranium concentration approximate as 8:2:1 in the cell wall organelle and cytosol fractions of roots of S. punctata. OM SEM and EDS showed after 5-200 mg/L U treatment 4-24 h, some intracellular fluid released from the root cells, after 100 mg/L U treatment 48 h, the particles including 35% Fe (wt%) and other organic matters such as EPS released from the cells, most of the uranium bound onto the root surface and contacted with phosphorus ligands and formed as nano-scales U-P lamellar crystal, similar crystal has been found in the cell wall and organelle fractions after 50 mg/L U treatment 120 h. FTIR and XPS analyses result indicates the uranium changed the band position and shapes of phosphate group, and the region of characteristic peak belongs to U(VI) and U(IV) were also observed.

Uranium from German Nuclear Power Projects of the 1940s--A Nuclear Forensic Investigation.

PubMed

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-11-02

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel. Through measurement of the (230)Th/(234)U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the (87)Sr/(86)Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of (236)U and (239)Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eppich, G.; Kips, R.; Lindvall, R.

The CUP-2 uranium ore concentrate (UOC) standard reference material, a powder, was produced at the Blind River uranium refinery of Eldorado Resources Ltd. in Canada in 1986. This material was produced as part of a joint effort by the Canadian Certified Reference Materials Project and the Canadian Uranium Producers Metallurgical Committee to develop a certified reference material for uranium concentration and the concentration of several impurity constituents. This standard was developed to satisfy the requirements of the UOC mining and milling industry, and was characterized with this purpose in mind. To produce CUP-2, approximately 25 kg of UOC derived frommore » the Blind River uranium refinery was blended, homogenized, and assessed for homogeneity by X-ray fluorescence (XRF) analysis. The homogenized material was then packaged into bottles, containing 50 g of material each, and distributed for analysis to laboratories in 1986. The CUP-2 UOC standard was characterized by an interlaboratory analysis program involving eight member laboratories, six commercial laboratories, and three additional volunteer laboratories. Each laboratory provided five replicate results on up to 17 analytes, including total uranium concentration, and moisture content. The selection of analytical technique was left to each participating laboratory. Uranium was reported on an “as-received” basis; all other analytes (besides moisture content) were reported on a “dry-weight” basis. A bottle of 25g of CUP-2 UOC standard as described above was purchased by LLNL and characterized by the LLNL Nuclear Forensics Group. Non-destructive and destructive analytical techniques were applied to the UOC sample. Information obtained from short-term techniques such as photography, gamma spectrometry, and scanning electron microscopy were used to guide the performance of longer-term techniques such as ICP-MS. Some techniques, such as XRF and ICP-MS, provided complementary types of data. The results indicate that the CUP-2 standard has a natural isotopic ratio, and does not appear to have been isotopically enriched or depleted in any way, and was not contaminated by a source of uranium with a non-natural isotopic composition. Furthermore, the lack of 233U and 236U above the instrumental detection limit indicates that this sample was not exposed to a neutron flux, which would have generated one or both of these isotopes in measurable concentrations.« less

Influence of microwaves on the leaching kinetics of uraninite from a low grade ore in dilute sulfuric acid.

PubMed

Madakkaruppan, V; Pius, Anitha; T, Sreenivas; Giri, Nitai; Sarbajna, Chanchal

2016-08-05

This paper describes a study on microwave assisted leaching of uranium from a low-grade ore of Indian origin. The host rock for uranium mineralization is chlorite-biotite-muscovite-quartzo-feldspathic schist. The dominant presence of siliceous minerals determined leaching of uranium values in sulfuric acid medium under oxidizing conditions. Process parametric studies like the effect of sulfuric acid concentration (0.12-0.50M), redox potential (400-500mV), particle size (600-300μm) and temperature (35°-95°C) indicated that microwave assisted leaching is more efficient in terms of overall uranium dissolution, kinetics and provide relatively less impurities (Si, Al, Mg and Fe) in the leach liquor compared to conventional conductive leaching. The kinetics of leaching followed shrinking core model with product layer diffusion as controlling mechanism. Copyright © 2016 Elsevier B.V. All rights reserved.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M

2014-01-01

The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making themore » gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.« less

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

Exopolysaccharide produced by Enterobacter sp. YG4 reduces uranium induced nephrotoxicity.

PubMed

K, Nagaraj; Devasya, Rekha Punchapady; Bhagwath, Arun Ananthapadmanabha

2016-01-01

Uranium nephrotoxicity is a health concern with very few treatment options. Bacterial exopolysaccharides (EPS) possess multiple biological activities and appear as prospective candidates for treating uranium nephrotoxicity. This study focuses on the ability of an EPS produced by a bacterial strain Enterobacter sp. YG4 to reduce uranium nephrotoxicity in vivo. This bacterium was isolated from the gut contents of a slug Laevicaulis alte (Férussac). Based on the aniline blue staining reaction and infrared spectral analysis, the EPS was identified as β-glucan and its molecular weight was 11.99×10(6)Da. The EPS showed hydroxyl radical scavenging ability and total antioxidant capacity in vitro. To assess the protection provided by the EPS against uranium nephrotoxicity, a single dose of 2mg/kg uranyl nitrate was injected intraperitoneally to albino Wistar rats. As intervention, the EPS was administered orally (100mg/kg/day) for 4 consecutive days. The rats were sacrificed on the fifth day and analyses were conducted. Increased serum creatinine and urea nitrogen levels and histopathological alterations in kidneys were observed in uranyl nitrate treated animals. All these alterations were reduced with the administration of Enterobacter sp. YG4 EPS, emphasizing a novel approach in treating uranium nephrotoxicity. Copyright © 2015 Elsevier B.V. All rights reserved.

Uranium in bone: metabolic and autoradiographic studies in the rat.

PubMed

Priest, N D; Howells, G R; Green, D; Haines, J W

1982-03-01

The distribution and retention of intravenously injected hexavalent uranium-233 in the skeleton of the female rat has been investigated using a variety of autoradiographic and radiochemical techniques. These showed that approximately one third of the injected uranium is deposited in the skeleton where it is retained with an initial biological half-time of approximately 40 days. The studies also showed that: 1 Uranium is initially deposited onto all types of bone surface, but preferentially onto those that are accreting. 2 Uranium is deposited in the calcifying zones of skeletal cartilage. 3 Bone accretion results in the burial of surface deposits of uranium. 4 Bone resorption causes the removal of uranium from surfaces. 5 Resorbed uranium is not retained by osteoclasts and macrophages in the bone marrow. 6 Uranium removed from bone surfaces enters the bloodstream where most is either redeposited in bone or excreted via the kidneys. 7 The recycling of resorbed uranium within the skeleton tends to produce a uniform level of uranium contamination throughout mineralized bone. These results are taken to indicate that uranium deposition in bone shares characteristics in common with both the 'volume-seeking radionuclides' typified by the alkaline earth elements and with the 'bone surface-seeking radionuclides' typified by plutonium.

Depleted and enriched uranium exposure quantified in former factory workers and local residents of NL Industries, Colonie, NY USA.

PubMed

Arnason, John G; Pellegri, Christine N; Moore, June L; Lewis-Michl, Elizabeth L; Parsons, Patrick J

2016-10-01

Between 1958 and 1982, NL Industries manufactured components of enriched (EU) and depleted uranium (DU) at a factory in Colonie NY, USA. More than 5 metric tons of DU was deposited as microscopic DU oxide particles on the plant site and surrounding residential community. A prior study involving a small number of individuals (n=23) indicated some residents were exposed to DU and former workers to both DU and EU, most probably through inhalation of aerosol particles. Our aim was to measure total uranium [U] and the uranium isotope ratios: (234)U/(238)U; (235)U/(238)U; and (236)U/(238)U, in the urine of a cohort of former workers and nearby residents of the NLI factory, to characterize individual exposure to natural uranium (NU), DU, and EU more than 3 decades after production ceased. We conducted a biomonitoring study in a larger cohort of 32 former workers and 99 residents, who may have been exposed during its period of operation, by measuring Total U, NU, DU, and EU in urine using Sector Field Inductively Coupled Plasma - Mass Spectrometry (SF-ICP-MS). Among workers, 84% were exposed to DU, 9% to EU and DU, and 6% to natural uranium (NU) only. For those exposed to DU, urinary isotopic and [U] compositions result from binary mixing of NU and the DU plant feedstock. Among residents, 8% show evidence of DU exposure, whereas none shows evidence of EU exposure. For residents, the [U] geometric mean is significantly below the value reported for NHANES. There is no significant difference in [U] between exposed and unexposed residents, suggesting that [U] alone is not a reliable indicator of exposure to DU in this group. Ninety four percent of workers tested showed evidence of exposure to DU, EU or both, and were still excreting DU and EU decades after leaving the workforce. The study demonstrates the advantage of measuring multiple isotopic ratios (e.g., (236)U/(238)U and (235)U/(238)U) over a single ratio ((235)U/(238)U) in determining sources of uranium exposure. Copyright © 2016 Elsevier Inc. All rights reserved.

Global Uranium And Thorium Resources: Are They Adequate To Satisfy Demand Over The Next Half Century?

NASA Astrophysics Data System (ADS)

Lambert, I. B.

2012-04-01

This presentation will consider the adequacy of global uranium and thorium resources to meet realistic nuclear power demand scenarios over the next half century. It is presented on behalf of, and based on evaluations by, the Uranium Group - a joint initiative of the OECD Nuclear Energy Agency and the International Atomic Energy Agency, of which the author is a Vice Chair. The Uranium Group produces a biennial report on Uranium Resources, Production and Demand based on information from some 40 countries involved in the nuclear fuel cycle, which also briefly reviews thorium resources. Uranium: In 2008, world production of uranium amounted to almost 44,000 tonnes (tU). This supplied approximately three-quarters of world reactor requirements (approx. 59,000 tU), the remainder being met by previously mined uranium (so-called secondary sources). Information on availability of secondary sources - which include uranium from excess inventories, dismantling nuclear warheads, tails and spent fuel reprocessing - is incomplete, but such sources are expected to decrease in market importance after 2013. In 2008, the total world Reasonably Assured plus Inferred Resources of uranium (recoverable at less than 130/kgU) amounted to 5.4 million tonnes. In addition, it is clear that there are vast amounts of uranium recoverable at higher costs in known deposits, plus many as yet undiscovered deposits. The Uranium Group has concluded that the uranium resource base is more than adequate to meet projected high-case requirements for nuclear power for at least half a century. This conclusion does not assume increasing replacement of uranium by fuels from reprocessing current reactor wastes, or by thorium, nor greater reactor efficiencies, which are likely to ameliorate future uranium demand. However, progressively increasing quantities of uranium will need to be mined, against a backdrop of the relatively small number of producing facilities around the world, geopolitical uncertainties and strong opposition to growth of nuclear power in a number of quarters - it is vital that the market provides incentives for exploration and development of environmentally sustainable mining operations. Thorium: World Reasonably Assured plus Inferred Resources of thorium are estimated at over 2.2 million tonnes, in hard rock and heavy mineral sand deposits. At least double this amount is considered to occur in as yet undiscovered thorium deposits. Currently, demand for thorium is insignificant, but even a major shift to thorium-fueled reactors would not make significant inroads into the huge resource base over the next half century.

TOF-SIMS for Rapid Nuclear Forensics Evaluation of Uranium Oxide Particles

DTIC Science & Technology

2011-03-01

Fraction U-238 nU U metal CRM 112-A NBL Metal Assay and Isotopic .000052458 .0072017 --- .9927458 nUO2 UO2 --- NBL Commercial material...0 .992745 dU U metal CRM 115 NBL Uranium Assay .0000076 .0020291 .0000322 .9979311 dUO2 UO2 --- IBI Labs Commercial material --- .002- .0035...U500* U3O8 CRM U500 NBL Isotopic .005181 .49696 .000755 .49711 U900* U3O8 CRM U900 NBL Isotopic .007777 .90196 .003327 .08693 *Sample

New Technique for Speciation of Uranium in Sediments Following Acetate-Stimulated Bioremediation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

2011-06-22

Acetate-stimulated bioremediation is a promising new technique for sequestering toxic uranium contamination from groundwater. The speciation of uranium in sediments after such bioremediation attempts remains unknown as a result of low uranium concentration, and is important to analyzing the stability of sequestered uranium. A new technique was developed for investigating the oxidation state and local molecular structure of uranium from field site sediments using X-Ray Absorption Spectroscopy (XAS), and was implemented at the site of a former uranium mill in Rifle, CO. Glass columns filled with bioactive Rifle sediments were deployed in wells in the contaminated Rifle aquifer and amendedmore » with a hexavalent uranium (U(VI)) stock solution to increase uranium concentration while maintaining field conditions. This sediment was harvested and XAS was utilized to analyze the oxidation state and local molecular structure of the uranium in sediment samples. Extended X-Ray Absorption Fine Structure (EXAFS) data was collected and compared to known uranium spectra to determine the local molecular structure of the uranium in the sediment. Fitting was used to determine that the field site sediments did not contain uraninite (UO{sub 2}), indicating that models based on bioreduction using pure bacterial cultures are not accurate for bioremediation in the field. Stability tests on the monomeric tetravalent uranium (U(IV)) produced by bioremediation are needed in order to assess the efficacy of acetate-stimulation bioremediation.« less

In situ spectroscopy and spectroelectrochemistry of uranium in high-temperature alkali chloride molten salts.

PubMed

Polovov, Ilya B; Volkovich, Vladimir A; Charnock, John M; Kralj, Brett; Lewin, Robert G; Kinoshita, Hajime; May, Iain; Sharrad, Clint A

2008-09-01

Soluble uranium chloride species, in the oxidation states of III+, IV+, V+, and VI+, have been chemically generated in high-temperature alkali chloride melts. These reactions were monitored by in situ electronic absorption spectroscopy. In situ X-ray absorption spectroscopy of uranium(VI) in a molten LiCl-KCl eutectic was used to determine the immediate coordination environment about the uranium. The dominant species in the melt was [UO 2Cl 4] (2-). Further analysis of the extended X-ray absorption fine structure data and Raman spectroscopy of the melts quenched back to room temperature indicated the possibility of ordering beyond the first coordination sphere of [UO 2Cl 4] (2-). The electrolytic generation of uranium(III) in a molten LiCl-KCl eutectic was also investigated. Anodic dissolution of uranium metal was found to be more efficient at producing uranium(III) in high-temperature melts than the cathodic reduction of uranium(IV). These high-temperature electrolytic processes were studied by in situ electronic absorption spectroelectrochemistry, and we have also developed in situ X-ray absorption spectroelectrochemistry techniques to probe both the uranium oxidation state and the uranium coordination environment in these melts.

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.

2016-11-01

The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performedmore » to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.« less

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

SINTERING METAL OXIDES

DOEpatents

Roake, W.E.

1960-09-13

A process is given for producing uranium dioxide material of great density by preparing a compacted mixture of uranium dioxide and from 1 to 3 wt.% of calcium hydride, heating the mixture to at least 675 deg C for decomposition of the hydride and then for sintering, preferably in a vacuum, at from 1550 to 2000 deg C. Calcium metal is formed, some uranium is reduced by the calcium to the metal and a product of high density is obtained.

Limiting Regret: Building the Army We Will Need

DTIC Science & Technology

2015-08-18

Recently, U.S. and Chinese experts have estimated that the North Koreans may be able to produce enough fissionable plutonium and uranium to build up...long-range missiles, but their recently revealed ability to separate uranium could give them the ability to build gun-assembled fission weapons similar...weapons programs and living up to their international obligations.” 36North Korea has had a uranium enrichment capacity since at least November 2010

FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

DOEpatents

Roake, W.E.

1958-08-19

A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

Iran’s Reemergence as a Major Player in Global Security

DTIC Science & Technology

2013-05-21

economic sanctions levied against the Islamic Republic. Iran continues to deny International Atomic Energy Agency inspectors’ access to possible uranium ...build nuclear weapons.”55 Mr. Clapper went on to say that “Iran’s technical advancement, particularly in uranium enrichment, strengthens our assessment...will to do so.”56 During the briefing, he made clear that Iran is technically capable of producing enough highly enriched uranium for a weapon

Method of making crack-free zirconium hydride

DOEpatents

Sullivan, Richard W.

1980-01-01

Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

Radionuclide tracers of sediment-water interactions on the Amazon shelf

NASA Astrophysics Data System (ADS)

Moore, Willard S.; DeMaster, David J.; Smoak, Joseph M.; McKee, Brent A.; Swarzenski, Peter W.

1996-04-01

A comprehensive study of a variety of radionuclide tracers has been coupled with other geochemical investigations and with sedimentary and physical oceanographic measurements to elucidate processes and their characteristic time scales at the mouth of the Amazon River. This two-year field study on the Amazon continental shelf involved four cruises designed to provide information during different stages of the river hydrograph. Although the cruises were coordinated with river stage, other physical variables including spring-neap tidal stages, the flow of the North Brazil Current and trade-wind stress caused important effects on the shelf environment. Partitioning of uranium among dissolved, colloidal and particulate phases was investigated during AmasSeds. A detailed examination of uranium water-column behavior during low river discharge found that most (89%) of the uranium near the Amazon River mouth was associated with the particulate phase and that most (92%) of the riverine dissolved-phase uranium was in the colloidal size fraction (0.001-0.4 μm). A non-conservative uranium/salinity distribution was observed for dissolved uranium, indicating large-scale uranium removal from surface waters with salinities less than 20 ppt. Colloidal uranium was non-conservative across the entire salinity regime, exhibiting removal of colloidal uranium from waters with salinities less than 12 ppt and a significant input at higher salinities. A short-lived particle-reactive tracer, 234Th, was used to evaluate the rates of particle scavenging on the shelf. Suspended-sediment concentrations respond to each turn of the tide, thus limiting the time available for equilibrium to be established between the particles and the tracers. Experiments demonstrated that on the Amazon shelf the partitioning and distribution of trace elements are governed by particle dynamics (particle residence times < sorption times). The high suspended load, including fluid muds, retards the incorporation of adsorbed 234Th into the seabed. Once scavenged, 234Th remains part of the suspended-sediment and fluid-mud inventory for periods of at least 4-8 weeks. Another particle-reactive tracer, 210Pb, was used to evaluate the potential supply of reactive metals from offshore waters to the shelf. As open-ocean waters move into the Amazon mixing zone, in response to the estuarine-like circulation, they lose 210Ph through scavenging processes associated with delta formation. This oceanic input of 210Pb dominates other inputs to the Amazon shelf system. Based on 210Pb analyses from more than 40 ☐ and kasten cores, the flux of water moving shoreward and depositing 210Pb in the sediments was calculated to be on the order of 6 × 10 161y -1 ˜10 times the riverine flux from the Amazon. The distribution of 210Pb in the sediments suggests that if particle-reactive species (such as certain trace metals) are released in dissolved form on the shelf, they will be scavenged quickly in this turbid environment, with the largest inventories occurring in the foreset beds (although the highest concentrations occur in the bottomset beds). The large landward flow of water indicates that if particle-reactive species are released in the western equatorial Atlantic via aerosol transport or other mechanisms, there is a good chance that a sizeable portion will be buried in the Amazon delta. Fluxes of radium isotopes, 226Ra, 228Ra and 224Ra, from the bottom sediments were used to evaluate sediment resuspension across the shelf. The average flux of 226Ra from the Amazon shelf balanced the annual desorption of 226Ra from river-derived sediments; however, departures between the 226Ra flux and sediment necessary to support the flux occurred for different sampling periods. During falling and low discharge, less sediment entered the system than was required to support the sedimentary desorption 226Ra flux. During rising and high discharge, more sediment entered than was necessary to sustain the 226Ra flux. Considerable recycling of particles between the seabed and water column was required to wash away most of the desorbable 226Ra from the sediment. To support the sedimentary 228Ra flux, 3.4 × 10 16 g of sediment must be resuspended each year. Such mixing would resuspend the top 35 em of sediment over the entire shelf each year, or mix the surface layer (SL, the region of uniform sedimentary 210Pb depth profiles) to a depth of 90 cm annually. Such mixing would turn over the entire SL every 1-2 years. Desorption of sedimentary radium isotopes must be accompanied by the desorption of other exchangeable species including phosphate, cesium and cadmium. Correlations among phosphate and radium isotopes, infer that phosphate is being released from bottom sediments as the sediments are resuspended into the water column. As was the case for 226Ra, bottom sediment must be repeatedly suspended into the water column to release all of the exchangeable phosphate. The desorption of 224Ra from shelf sediments provides a tracer of the Amazon plume into the Atlantic. In July and September 1989, the ship crossed the Demerara retroflection 380 km from the Amazon shelf and excess 224Ra activities up to 8 dpm 100 kg -1 were measured. It was concluded that these waters were <5 days removed from the Amazon shelf. A sustained current of >80 em s -1 would be required to advect the water this distance in <5 days, an observation that agrees well with drogue releases during August 1989.

Efficacy of acetate-amended biostimulation for uranium sequestration: Combined analysis of sediment/groundwater geochemistry and bacterial community structure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Jie; Veeramani, Harish; Qafoku, Nikolla P.

Systematic flow-through column experiments were conducted using sediments and ground water collected from different subsurface localities at the U.S. Department of Energy's Integrated Field Research Challenge site in Rifle, Colorado. The principal purpose of this study is to gain a better understanding of the interactive effects of groundwater geochemistry, sediment mineralogy, and indigenous bacterial community structures on the efficacy of uranium removal from the groundwater with/without acetate amendment. Overall, we find that the subtle variations in the sediments' mineralogy, particle size, redox conditions, as well as contents of metal(loid) co-contaminants showed a pronounced effect on the associated bacterial population andmore » composition, which mainly determines the system's performance with respect to uranium removal. Positive relationship was identified between the abundance of dissimilatory sulfate-reduction genes (i.e., drsA), markers of sulfatereducing bacteria, and the sediments' propensity to sequester aqueous uranium. In contrast, no obvious connections were observed between the abundance of common iron-reducing bacteria, e.g., Geobacter spp., and the sediments' ability to sequester uranium. In the sediments with low bacterial biomass and the absence of sulfate-reducing conditions, abiotic adsorption onto mineral surfaces such as phyllosilicates likely played a relatively major role in the attenuation of aqueous uranium; however, in these scenarios, acetate amendment induced detectable rebounds in the effluent uranium concentrations. The results of this study suggest that reductive immobilization of uranium can be achieved under predominantly sulfate-reducing conditions, and provide insight into the integrated roles of various biogeochemical components in long-term uranium sequestration.« less

Laser removal of loose uranium compound contamination from metal surfaces

NASA Astrophysics Data System (ADS)

Roberts, D. E.; Modise, T. S.

2007-04-01

Pulsed laser removal of surface contamination of uranyl nitrate and uranium dioxide from stainless steel has been studied. Most of the loosely bound contamination has been removed at fluence levels below 0.5 J cm -2, leaving about 5% fixed contamination for uranyl nitrate and 15% for uranium dioxide. Both alpha and beta activities are then sufficiently low that contaminated objects can be taken out of a restricted radiation area for re-use. The ratio of beta to alpha activity is found to be a function of particle size and changes during laser removal. In a separate experiment using technetium-99m, the collection of removed radioactivity in the filter was studied and an inventory made of removed and collected contamination.

DIRECT INGOT PROCESS FOR PRODUCING URANIUM

DOEpatents

Leaders, W.M.; Knecht, W.S.

1960-11-15

A process is given in which uranium tetrafluoride is reduced to the metal with magnesium and in the same step the uranium metal formed is cast into an ingot. For this purpose a mold is arranged under and connected with the reaction bomb, and both are filled with the reaction mixture. The entire mixture is first heated to just below reaction temperature, and thereafter heating is restricted to the mixture in the mold. The reaction starts in the mold whereby heat is released which brings the rest of the mixture to reaction temperature. Pure uranium metal settles in the mold while the magnesium fluoride slag floats on top of it. After cooling, the uranium is separated from the slag by mechanical means.

Compatibility of buffered uranium carbides with tungsten.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1971-01-01

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

Different biosorption mechanisms of Uranium(VI) by live and heat-killed Saccharomyces cerevisiae under environmentally relevant conditions.

PubMed

Wang, Tieshan; Zheng, Xinyan; Wang, Xiaoyu; Lu, Xia; Shen, Yanghao

2017-02-01

Uranium adsorption mechanisms of live and heat-killed Saccharomyces cerevisiae in different pH values and biomass concentrations were studied under environmentally relevant conditions. Compared with live cells, the adsorption capacity of heat-killed cells is almost one order of magnitude higher in low biomass concentration and highly acidic pH conditions. To explore the mesoscopic surface interactions between uranium and cells, the characteristic of uranium deposition was investigated by SEM-EDX, XPS and FTIR. Biosorption process of live cells was considered to be metabolism-dependent. Under stimulation by uranyl ions, live cells could gradually release phosphorus and reduce uranium from U(VI) to U(IV) to alleviate uranium toxicity. The uranyl-phosphate complexes were formed in scale-like shapes on cell surface. The metabolic detoxification mechanisms such as reduction and "self-protection" are of significance to the migration of radionuclides. In the metabolism-independent biosorption process of heat-killed cells: the cells cytomembrane was damaged by autoclaving which led to the free diffusion of phosphorous from intracellular, and the rough surface and nano-holes indicated that the dead cells provided larger contact area to precipitate U(VI) as spherical nano-particles. The high biosorption capacity of heat-killed cells makes it become a suitable biological adsorbent for uranium removal. Copyright © 2016 Elsevier Ltd. All rights reserved.

Homologation and functionalization of carbon monoxide by a recyclable uranium complex.

PubMed

Gardner, Benedict M; Stewart, John C; Davis, Adrienne L; McMaster, Jonathan; Lewis, William; Blake, Alexander J; Liddle, Stephen T

2012-06-12

Carbon monoxide (CO) is in principle an excellent resource from which to produce industrial hydrocarbon feedstocks as alternatives to crude oil; however, CO has proven remarkably resistant to selective homologation, and the few complexes that can effect this transformation cannot be recycled because liberation of the homologated product destroys the complexes or they are substitutionally inert. Here, we show that under mild conditions a simple triamidoamine uranium(III) complex can reductively homologate CO and be recycled for reuse. Following treatment with organosilyl halides, bis(organosiloxy)acetylenes, which readily convert to furanones, are produced, and this was confirmed by the use of isotopically (13)C-labeled CO. The precursor to the triamido uranium(III) complex is formed concomitantly. These findings establish that, under appropriate conditions, uranium(III) can mediate a complete synthetic cycle for the homologation of CO to higher derivatives. This work may prove useful in spurring wider efforts in CO homologation, and the simplicity of this system suggests that catalytic CO functionalization may soon be within reach.

β-decay Rates for Exotic Nuclei and r-process Nucleosynthesis up to Thorium and Uranium

NASA Astrophysics Data System (ADS)

Suzuki, Toshio; Shibagaki, Shota; Yoshida, Takashi; Kajino, Toshitaka; Otsuka, Takaharu

2018-06-01

Beta-decay rates for exotic nuclei with neutron magic number of N = 126 relevant to r-process nucleosynthesis are studied up to Z = 78 by shell-model calculations. The half-lives for the waiting-point nuclei obtained, which are short compared to a standard finite-range-droplet model, are used to study r-process nucleosynthesis in core-collapse supernova (CCSN) explosions and binary neutron star mergers. The element abundances are obtained up to the third peak as well as beyond the peak region up to thorium and uranium. The position of the third peak is found to be shifted toward a higher mass region in both CCSN explosions and neutron star mergers. We find that thorium and uranium elements are produced more with the shorter shell-model half-lives and their abundances come close to the observed values in CCSN explosions. In the case of binary neutron star mergers, thorium and uranium are produced consistently with the observed values independent of the half-lives.

Homologation and functionalization of carbon monoxide by a recyclable uranium complex

PubMed Central

Gardner, Benedict M.; Stewart, John C.; Davis, Adrienne L.; McMaster, Jonathan; Lewis, William; Blake, Alexander J.; Liddle, Stephen T.

2012-01-01

Carbon monoxide (CO) is in principle an excellent resource from which to produce industrial hydrocarbon feedstocks as alternatives to crude oil; however, CO has proven remarkably resistant to selective homologation, and the few complexes that can effect this transformation cannot be recycled because liberation of the homologated product destroys the complexes or they are substitutionally inert. Here, we show that under mild conditions a simple triamidoamine uranium(III) complex can reductively homologate CO and be recycled for reuse. Following treatment with organosilyl halides, bis(organosiloxy)acetylenes, which readily convert to furanones, are produced, and this was confirmed by the use of isotopically 13C-labeled CO. The precursor to the triamido uranium(III) complex is formed concomitantly. These findings establish that, under appropriate conditions, uranium(III) can mediate a complete synthetic cycle for the homologation of CO to higher derivatives. This work may prove useful in spurring wider efforts in CO homologation, and the simplicity of this system suggests that catalytic CO functionalization may soon be within reach. PMID:22652572

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

PubMed

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fincke, J.R.; Swank, W.D.; Haggard, D.C.

This paper describes the experimental demonstration of a process for the direct plasma reduction of depleted uranium hexafluoride to uranium metal. The process exploits the large departures from equilibrium that can be achieved in the rapid supersonic expansion of a totally dissociated and partially ionized mixture of UF{sub 6}, Ar, He, and H{sub 2}. The process is based on the rapid condensation of subcooled uranium vapor and the relatively slow rate of back reaction between metallic uranium and HF to F{sub 2} to reform stable fluorides. The high translational velocities and rapid cooling result in an overpopulation of atomic hydrogenmore » which persists throughout the expansion process. Atomic hydrogen shifts the equilibrium composition by inhibiting the reformation of uranium-fluorine compounds. This process has the potential to reduce the cost of reducing UF{sub 6} to uranium metal with the added benefit of being a virtually waste free process. The dry HF produced is a commodity which has industrial value.« less

Identifying anthropogenic uranium compounds using soft X-ray near-edge absorption spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, Jesse D.; Bowden, Mark; Tom Resch, C.

2017-01-01

Uranium ores mined for industrial use are typically acid-leached to produce yellowcake and then converted into uranium halides for enrichment and purification. These anthropogenic chemical forms of uranium are distinct from their mineral counterparts. The purpose of this study is to use soft X-ray absorption spectroscopy to characterize several common anthropogenic uranium compounds important to the nuclear fuel cycle. Non-destructive chemical analyses of these compounds is important for process and environmental monitoring and X-ray absorption techniques have several advantages in this regard, including element-specificity, chemical sensitivity, and high spectral resolution. Oxygen K-edge spectra were collected for uranyl nitrate, uranyl fluoride,more » and uranyl chloride, and fluorine K-edge spectra were collected for uranyl fluoride and uranium tetrafluoride. Interpretation of the data is aided by comparisons to calculated spectra. These compounds have unique spectral signatures that can be used to identify unknown samples.« less

Production of fissioning uranium plasma to approximate gas-core reactor conditions

NASA Technical Reports Server (NTRS)

Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

1974-01-01

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

DOEpatents

Vijayan, S.; Wong, C.F.; Buckley, L.P.

1994-11-22

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

DOEpatents

Vijayan, Sivaraman; Wong, Chi F.; Buckley, Leo P.

1994-01-01

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved.

The SPES High Power ISOL production target

NASA Astrophysics Data System (ADS)

Andrighetto, A.; Corradetti, S.; Ballan, M.; Borgna, F.; Manzolaro, M.; Scarpa, D.; Monetti, A.; Rossignoli, M.; Silingardi, R.; Mozzi, A.; Vivian, G.; Boratto, E.; De Ruvo, L.; Sattin, N.; Meneghetti, G.; Oboe, R.; Guerzoni, M.; Margotti, A.; Ferrari, M.; Zenoni, A.; Prete, G.

2016-11-01

SPES (Selective Production of Exotic Species) is a facility under construction at INFN-LNL (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro), aimed to produce intense neutron-rich radioactive ion beams (RIBs). These will be obtained using the ISOL (Isotope Separation On-Line) method, bombarding a uranium carbide target with a proton beam of 40MeV energy and currents up to 200μA. The target configuration was designed to obtain a high number of fissions, up to 1013 per second, low power deposition and fast release of the produced isotopes. The exotic isotopes generated in the target are ionized, mass separated and re-accelerated by the ALPI superconducting LINAC at energies of 10AMeV and higher, for masses in the region of A = 130 amu , with an expected rate on the secondary target up to 109 particles per second. In this work, recent results on the R&D activities regarding the SPES RIB production target-ion source system are reported.

10 CFR 39.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... neutron generator tube to produce neutrons for use in well logging applications. Uranium sinker bar means a weight containing depleted uranium used to pull a logging tool toward the bottom of a well. Well... drilled holes for the purpose of oil, gas, mineral, groundwater, or geological exploration. Well logging...

10 CFR 39.2 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... neutron generator tube to produce neutrons for use in well logging applications. Uranium sinker bar means a weight containing depleted uranium used to pull a logging tool toward the bottom of a well. Well... drilled holes for the purpose of oil, gas, mineral, groundwater, or geological exploration. Well logging...

10 CFR 39.2 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... neutron generator tube to produce neutrons for use in well logging applications. Uranium sinker bar means a weight containing depleted uranium used to pull a logging tool toward the bottom of a well. Well... drilled holes for the purpose of oil, gas, mineral, groundwater, or geological exploration. Well logging...

10 CFR 39.2 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... neutron generator tube to produce neutrons for use in well logging applications. Uranium sinker bar means a weight containing depleted uranium used to pull a logging tool toward the bottom of a well. Well... drilled holes for the purpose of oil, gas, mineral, groundwater, or geological exploration. Well logging...

10 CFR 39.2 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... neutron generator tube to produce neutrons for use in well logging applications. Uranium sinker bar means a weight containing depleted uranium used to pull a logging tool toward the bottom of a well. Well... drilled holes for the purpose of oil, gas, mineral, groundwater, or geological exploration. Well logging...

9. VIEW OF FOUNDRY FURNACE, DEPLETED URANIUM INGOTS, BERYLLIUM INGOTS, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. VIEW OF FOUNDRY FURNACE, DEPLETED URANIUM INGOTS, BERYLLIUM INGOTS, AND ALUMINUM SHAPES WERE PRODUCED IN THE FOUNDRY. (10/30/56) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

Temperature Dependence of Uranium and Vanadium Adsorption on Amidoxime-Based Adsorbents in Natural Seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuo, Li-Jung; Gill, Gary A.; Tsouris, Costas

Recent advances in the development of amidoxime-based adsorbents have made it highly promising for seawater uranium extraction. However, there is a great need to understand the influence of temperature on the uranium sequestration performance of the adsorbents in natural seawater. Here in this work, the apparent enthalpy and entropy of the sorption of uranium (VI) and vanadium (V) with amidoxime-based adsorbents were determined in natural seawater tests at 8, 20, and 31 °C that cover a broad range of ambient seawater temperature. The sorption of U was highly endothermic, producing apparent enthalpies of 57 ± 6.0 and 59 ± 11more » kJ mol -1 and apparent entropies of 314 ± 21 and 320 ± 36 J K-1 mol -1, respectively, for two adsorbent formulations. In contrast, the sorption of V showed a much smaller temperature sensitivity, producing apparent enthalpies of 6.1 ± 5.9 and -11 ± 5.7 kJ mol -1 and apparent entropies of 164 ± 20 and 103 ± 19 J K -1 mol -1, respectively. This new thermodynamic information suggests that amidoxime-based adsorbents will deliver significantly increased U adsorption capacities and improved selectivity in warmer waters. A separate field study of seawater uranium adsorption conducted in a warm seawater site (Miami, FL, USA) confirm the observed strong temperature effect on seawater uranium mining. Lastly, this strong temperature dependence demonstrates that the warmer the seawater where the amidoxime-based adsorbents are deployed the greater the yield for seawater uranium extraction.« less

Temperature Dependence of Uranium and Vanadium Adsorption on Amidoxime-Based Adsorbents in Natural Seawater

DOE PAGES

Kuo, Li-Jung; Gill, Gary A.; Tsouris, Costas; ...

2018-01-16

Recent advances in the development of amidoxime-based adsorbents have made it highly promising for seawater uranium extraction. However, there is a great need to understand the influence of temperature on the uranium sequestration performance of the adsorbents in natural seawater. Here in this work, the apparent enthalpy and entropy of the sorption of uranium (VI) and vanadium (V) with amidoxime-based adsorbents were determined in natural seawater tests at 8, 20, and 31 °C that cover a broad range of ambient seawater temperature. The sorption of U was highly endothermic, producing apparent enthalpies of 57 ± 6.0 and 59 ± 11more » kJ mol -1 and apparent entropies of 314 ± 21 and 320 ± 36 J K-1 mol -1, respectively, for two adsorbent formulations. In contrast, the sorption of V showed a much smaller temperature sensitivity, producing apparent enthalpies of 6.1 ± 5.9 and -11 ± 5.7 kJ mol -1 and apparent entropies of 164 ± 20 and 103 ± 19 J K -1 mol -1, respectively. This new thermodynamic information suggests that amidoxime-based adsorbents will deliver significantly increased U adsorption capacities and improved selectivity in warmer waters. A separate field study of seawater uranium adsorption conducted in a warm seawater site (Miami, FL, USA) confirm the observed strong temperature effect on seawater uranium mining. Lastly, this strong temperature dependence demonstrates that the warmer the seawater where the amidoxime-based adsorbents are deployed the greater the yield for seawater uranium extraction.« less

Uranium extraction by complexation with siderophores

NASA Astrophysics Data System (ADS)

Bahamonde Castro, Cristina

One of the major concerns of energy production is the environmental impact associated with the extraction of natural resources. Nuclear energy fuel is obtained from uranium, an abundant and naturally occurring element in the environment, but the currently used techniques for uranium extraction leave either a significant fingerprint (open pit mines) or a chemical residue that alters the pH of the environment (acid or alkali leaching). It is therefore clear that a new and greener approach to uranium extraction is needed. Bioleaching is one potential alternative. In bioleaching, complexants naturally produced from fungi or bacteria may be used to extract the uranium. In the following research, the siderophore enterobactin, which is naturally produced by bacteria to extract and solubilize iron from the environment, is evaluated to determine its potential for complexing with uranium. To determine whether enterobactin could be used for uranium extraction, its acid dissociation and its binding strength with the metal of interest must be determined. Due to the complexity of working with radioactive materials, lanthanides were used as analogs for uranium. In addition, polyprotic acids were used as structural and chemical analogs for the siderophore during method development. To evaluate the acid dissociation of enterobactin and the subsequent binding constants with lanthanides, three different analytical techniques were studied including: potentiometric titration, UltraViolet Visible (UV-Vis) spectrophotometry and Isothermal Titration Calorimetry (ITC). After evaluation of three techniques, a combination of ITC and potentiometric titrations was deemed to be the most viable way for studying the siderophore of interest. The results obtained from these studies corroborate the ideal pH range for enterobactin complexation to the lanthanide of interest and pave the way for determining the strength of complexation relative to other naturally occurring metals. Ultimately, this fundamental research enhances our current understanding of heavy metal complexation to naturally occurring complexants, which may enhance the metals mobility in the environment or potentially be used as a greener alternative in uranium extraction or remediation.

Caulobacter crescentus as a Whole-Cell Uranium Biosensor▿ †

PubMed Central

Hillson, Nathan J.; Hu, Ping; Andersen, Gary L.; Shapiro, Lucy

2007-01-01

We engineered a strain of the bacterium Caulobacter crescentus to fluoresce in the presence of micromolar levels of uranium at ambient temperatures when it is exposed to a hand-held UV lamp. Previous microarray experiments revealed that several Caulobacter genes are significantly upregulated in response to uranium but not in response to other heavy metals. We designated one of these genes urcA (for uranium response in caulobacter). We constructed a reporter that utilizes the urcA promoter to produce a UV-excitable green fluorescent protein in the presence of the uranyl cation, a soluble form of uranium. This reporter is specific for uranium and has little cross specificity for nitrate (<400 μM), lead (<150 μM), cadmium (<48 μM), or chromium (<41.6 μM). The uranium reporter construct was effective for discriminating contaminated groundwater samples (4.2 μM uranium) from uncontaminated groundwater samples (<0.1 μM uranium) collected at the Oak Ridge Field Research Center. In contrast to other uranium detection methodologies, the Caulobacter reporter strain can provide on-demand usability in the field; it requires minimal sample processing and no equipment other than a hand-held UV lamp, and it may be sprayed directly on soil, groundwater, or industrial surfaces. PMID:17905881

Uranium in surface soils: an easy-and-quick assay combining X-ray diffraction and fluorescence qualitative data

NASA Astrophysics Data System (ADS)

Figueiredo, M. O.; Silva, T. P.; Batista, M. J.; Leote, J.; Ferreira, M. L.; Limpo, V.

2009-04-01

Portugal has been a uranium-producer since the beginning of the last century. The uranium-rich area of Alto Alentejo, East-central Portugal, was identified more than fifty years ago [1]. Almost all the uranium-bearing mineralization occurs in schistose rocks of the contact metamorphic aureole produced by intrusion of the Hercynian monzonitic granite of Alto Alentejo into the pre-Ordovitian schist-greywacke complex forming deposits of vein and dissemination type. The Nisa uranium-reservoir, situated at the sharp border of a large and arch shaped granite pluton, was identified in 1957 [2] but its exploitation was considered economically impracticable until recently. However, its existence and the accumulated detritus of these prospect efforts are a concern for local populations [3]. A study of the near-surface soils close to the Nisa reservoir was therefore undertaken to assess the uranium retention by adsorption on clay components under the form of uranyl ions, [UO2]2+ [4-6] and its eventual release into the aquifer groundwater. As an attempt to very quickly appraise the presence of uranium in as-collected near-surface sediment samples a combination of laboratory X-ray techniques was designed: X-ray diffraction (XRD) to identify the mineral phases and roughly estimate its relative proportion plus X-ray fluorescence spectrometry in wavelength dispersive mode (XRF-WDS) to ascertain the presence of uranium and tentatively evaluate its content by comparison with selected chemical components of the soil. A description of the experimental methodology adopted for the implemented easy-and-quick uranium assay is presented. Obtained results compare quite well to the data of certified time-consuming analytical tests of uranium in those soil samples. [1] L. Pilar (1966) Conditions of formation of Nisa uranium deposit (in Portuguese). Comunic. Serv. Geol. Portugal, tomo L, 50-85. [2] C. Gonçalves & J.V. Teixeira Lopes (1971) Uranium deposit of Nisa: geological aspects of its discovery and valorisation (in Portuguese). Internal Rept., JEN, 20 pp. [3] http://www.naturtejo.com [4] J.A. Davis et al. (2006) Processes affecting transport of uranium in a suboxic aquifer. Phys. Chem. of the Earth 31, 548-555. [5] Y. Arai et al. (2007) Spectroscopic evidence for uranium bearing precipitates in Vadose zone sediments at the Hanford 300-Area site. Environ. Sci. Technol. 41, 4633-4639. [6] A. Kremleva, S. Krüger & N. Rösch (2008) Density functional model studies of uranyl adsorption on (001) surfaces of kaolinite. Langmuir 24, 9515-9524.

Triso coating development progress for uranium nitride kernels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jolly, Brian C.; Lindemer, Terrence; Terrani, Kurt A.

2015-08-01

In support of fully ceramic matrix (FCM) fuel development [1-2], coating development work is ongoing at the Oak Ridge National Laboratory (ORNL) to produce tri-structural isotropic (TRISO) coated fuel particles with UN kernels [3]. The nitride kernels are used to increase fissile density in these SiC-matrix fuel pellets with details described elsewhere [4]. The advanced gas reactor (AGR) program at ORNL used fluidized bed chemical vapor deposition (FBCVD) techniques for TRISO coating of UCO (two phase mixture of UO2 and UCx) kernels [5]. Similar techniques were employed for coating of the UN kernels, however significant changes in processing conditions weremore » required to maintain acceptable coating properties due to physical property and dimensional differences between the UCO and UN kernels (Table 1).« less

WNA's worldwide overview on front-end nuclear fuel cycle growth and health, safety and environmental issues.

PubMed

Saint-Pierre, Sylvain; Kidd, Steve

2011-01-01

This paper presents the WNA's worldwide nuclear industry overview on the anticipated growth of the front-end nuclear fuel cycle from uranium mining to conversion and enrichment, and on the related key health, safety, and environmental (HSE) issues and challenges. It also puts an emphasis on uranium mining in new producing countries with insufficiently developed regulatory regimes that pose greater HSE concerns. It introduces the new WNA policy on uranium mining: Sustaining Global Best Practices in Uranium Mining and Processing-Principles for Managing Radiation, Health and Safety and the Environment, which is an outgrowth of an International Atomic Energy Agency (IAEA) cooperation project that closely involved industry and governmental experts in uranium mining from around the world. Copyright © 2010 Health Physics Society

Long-term ecological effects of exposure to uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, W.C.; Miera, F.R. Jr.

1976-03-01

The consequences of releasing natural and depleted uranium to terrestrial ecosystems during development and testing of depleted uranium munitions were investigated. At Eglin Air Force Base, Florida, soil at various distances from armor plate target butts struck by depleted uranium penetrators was sampled. The upper 5 cm of soil at the target bases contained an average of 800 ppM of depleted uranium, about 30 times as much as soil at 5- to 10-cm depth, indicating some vertical movement of depleted uranium. Samples collected beyond about 20 m from the targets showed near-background natural uranium levels, about 1.3 +- 0.3 ..mu..g/gmore » or ppM. Two explosives-testing areas at the Los Alamos Scientific Laboratory (LASL) were selected because of their use history. E-F Site soil averaged 2400 ppM of uranium in the upper 5 cm and 1600 ppM at 5-10 cm. Lower Slobovia Site soil from two subplots averaged about 2.5 and 0.6 percent of the E-F Site concentrations. Important uranium concentration differences with depth and distance from detonation points were ascribed to the different explosive tests conducted in each area. E-F Site vegetation samples contained about 320 ppM of uranium in November 1974 and about 125 ppM in June 1975. Small mammals trapped in the study areas in November contained a maximum of 210 ppM of uranium in the gastrointestinal tract contents, 24 ppM in the pelt, and 4 ppM in the remaining carcass. In June, maximum concentrations were 110, 50, and 2 ppM in similar samples and 6 ppM in lungs. These data emphasized the importance of resuspension of respirable particles in the upper few millimeters of soil as a contamination mechanism for several components of the LASL ecosystem.« less

Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

NASA Astrophysics Data System (ADS)

McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

2009-11-01

This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

Efficacy of acetate-amended biostimulation for uranium sequestration: Combined analysis of sediment/groundwater geochemistry and bacterial community structure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Jie; Veeramani, Harish; Qafoku, Nikolla P.

Systematic flow-through column experiments were conducted using sediments and ground water collected from different subsurface localities at the U.S. Department of Energy’s Integrated Field Research Challenge site in Rifle, Colorado. The principal purpose of this study is to gain a better understanding of the interactive effects of groundwater geochemistry, sediment mineralogy, and indigenous bacterial community structures on the efficacy of uranium removal from the groundwater with/without acetate amendment. Overall, we find that the subtle variations in the sediments’ mineralogy, particle size, redox conditions, as well as contents of metal(loid) co-contaminants showed a pronounced effect on the associated bacterial population andmore » composition, which mainly determines the system’s performance with respect to uranium removal. Positive relationship was identified between the abundance of dissimilatory sulfate-reduction genes (i.e., drsA), markers of sulfate-reducing bacteria, and the sediments’ propensity to sequester aqueous uranium. In contrast, no obvious connections were observed between the abundance of common iron-reducing bacteria, e.g., Geobacter spp., and the sediments’ ability to sequester uranium. In the sediments with low bacterial biomass and the absence of sulfate-reducing conditions, abiotic adsorption onto mineral surfaces such as phyllosilicates likely played a relatively major role in the attenuation of aqueous uranium; however, in these scenarios, acetate amendment induced detectable rebounds in the effluent uranium concentrations. The results of this study suggest that reductive immobilization of uranium can be achieved under predominantly sulfate-reducing conditions, and provide insight into the integrated roles of various biogeochemical components in long-term uranium sequestration.« less

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

NASA Astrophysics Data System (ADS)

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.; Lavender, Curt A.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum (U10Mo) coupon was studied by plane-strain compression finite element modeling. As-cast U10Mo typically contains second phase particles such as uranium carbides (UC) and silicides along the grain boundaries. The volume fraction of UC is typically large, while the other phases can be redissolved in the matrix by certain heat treatments. The UC particle distribution is important due to its influence on the recrystallization processes (particle stimulated nucleation) that occur during annealing between rolling passes. Unfavorable particle distribution and fracture after rolling can affect the grain size and also influence the fuel performance in the reactor. A statistical method, i.e., the two-point correlation function (2PCF), was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis shows that the interparticle spacing shrinks along the normal direction (ND) and increases along the rolling direction (RD). The simulated particle distribution is very similar to that measured experimentally for similar rolling reductions. The magnitudes of major peaks of 2PCF along the ND decrease after large reduction. The locations of major peaks indicate the inter-stringer distances. Many more small peaks appear for the 2PCF along the RD, and this is related to the neighboring particles within stringers, which are along the RD.

PREPARATION OF URANIUM TRIOXIDE

DOEpatents

Buckingham, J.S.

1959-09-01

The production of uranium trioxide from aqueous solutions of uranyl nitrate is discussed. The uranium trioxide is produced by adding sulfur or a sulfur-containing compound, such as thiourea, sulfamic acid, sulfuric acid, and ammonium sulfate, to the uranyl solution in an amount of about 0.5% by weight of the uranyl nitrate hexahydrate, evaporating the solution to dryness, and calcining the dry residue. The trioxide obtained by this method furnished a dioxide with a considerably higher reactivity with hydrogen fluoride than a trioxide prepared without the sulfur additive.

222Rn emanation from uranium-glazed ceramics.

PubMed

Biagioni, R N; Sheets, R W

2001-02-01

Orange-red uranium-glazed dinnerware was found to emanate 222Rn to produce an average of less than 0.1 Bq 222Rn decay per piece, which should have no associated health risks. Comparison of 222Rn emanation to 226Ra in glazes (measured by alpha spectrometry) indicated inefficient emanation (<5%) of 222Rn, consistent with the low radon levels observed and with the presence of 222Rn progeny in glazes. These studies also showed that reagent grade uranium compounds may emanate measurable 222Rn.

Simulations of Ground-Water Flow and Particle Pathline Analysis in the Zone of Contribution of a Public-Supply Well in Modesto, Eastern San Joaquin Valley, California

USGS Publications Warehouse

Burow, Karen R.; Jurgens, Bryant C.; Kauffman, Leon J.; Phillips, Steven P.; Dalgish, Barbara A.; Shelton, Jennifer L.

2008-01-01

Shallow ground water in the eastern San Joaquin Valley is affected by high nitrate and uranium concentrations and frequent detections of pesticides and volatile organic compounds (VOC), as a result of ground-water development and intensive agricultural and urban land use. A single public-supply well was selected for intensive study to evaluate the dominant processes affecting the vulnerability of public-supply wells in the Modesto area. A network of 23 monitoring wells was installed, and water and sediment samples were collected within the approximate zone of contribution of the public-supply well, to support a detailed analysis of physical and chemical conditions and processes affecting the water chemistry in the well. A three-dimensional, steady-state local ground-water-flow and transport model was developed to evaluate the age of ground water reaching the well and to evaluate the vulnerability of the well to nonpoint source input of nitrate and uranium. Particle tracking was used to compute pathlines and advective travel times in the ground-water flow model. The simulated ages of particles reaching the public-supply well ranged from 9 to 30,000 years, with a median of 54 years. The age of the ground water contributed to the public-supply well increased with depth below the water table. Measured nitrate concentrations, derived primarily from agricultural fertilizer, were highest (17 milligrams per liter) in shallow ground water and decreased with depth to background concentrations of less than 2 milligrams per liter in the deepest wells. Because the movement of water is predominantly downward as a result of ground-water development, and because geochemical conditions are generally oxic, high nitrate concentrations in shallow ground water are expected to continue moving downward without significant attenuation. Simulated long-term nitrate concentrations indicate that concentrations have peaked and will decrease in the public-supply well during the next 100 years because of the low nitrate concentrations in recharge beneath the urban area and the increasing proportion of urban-derived ground water reaching the well. The apparent lag time between peak input concentrations and peak concentrations in the well is about 20 to 30 years. Measured uranium concentrations were also highest (45 micrograms per liter) in shallow ground water, and decreased with depth to background concentrations of about 0.5 microgram per liter. Naturally-occurring uranium adsorbed to aquifer sediments is mobilized by oxygen-rich, high-alkalinity water. Alkalinity increased in shallow ground water in response to agricultural development. As ground-water pumping increased in the 1940s and 1950s, this alkaline water moved downward through the ground-water flow system, mobilizing the uranium adsorbed to aquifer sediments. Ground water with high alkalinity and high uranium concentrations is expected to continue to move deeper in the system, resulting in increased uranium concentrations with depth in ground water. Because alkalinity (and correspondingly uranium) concentrations were high in shallow ground water beneath both the urban and the agricultural land, long-term uranium concentrations in the public-supply well are expected to increase as the proportion of uranium-affected water contributed to the well increases. Assuming that the alkalinity near the water table remains the same, the simulation of long-term alkalinity in the public-supply well indicates that uranium concentrations in the public-supply well will likely approach the maximum contaminant level; however, the time to reach this level is more than 100 years because of the significant proportion of old, unaffected water at depth that is contributed to the public-supply well.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-09-30

to supply uranium,” The Hindu, January 25, 2009; Kazakhstan might start uranium exports to India in 2009,” Panorama , February 6, 2009. “Chennai Daily...93-485 amended Section 123 d. to include agreements that covered reactors producing more than 5 MW thermal or special nuclear material connected

Atomic Mass and Nuclear Binding Energy for U-287 (Uranium)

NASA Astrophysics Data System (ADS)

Sukhoruchkin, S. I.; Soroko, Z. N.

This document is part of the Supplement containing the complete sets of data of Subvolume B `Nuclei with Z = 55 - 100' of Volume 22 `Nuclear Binding Energies and Atomic Masses' of Landolt-Börnstein - Group I `Elementary Particles, Nuclei and Atoms', and additionally including data for nuclei with Z = 101 - 130. It provides atomic mass, mass excess, nuclear binding energy, nucleon separation energies, Q-values, and nucleon residual interaction parameters for atomic nuclei of the isotope U-287 (Uranium, atomic number Z = 92, mass number A = 287).

Low-enriched uranium high-density target project. Compendium report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vandegrift, George; Brown, M. Alex; Jerden, James L.

2016-09-01

At present, most 99Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassemblymore » of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.« less

Uranium hydrogeochemical and stream sediment reconnaissance of the Newcastle NTMS Quadrangle, Wyoming, including concentrations of forty-two additional elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, S.J.; Sandoval, W.F.; Gallimore, D.L.

1980-06-01

Water and sediment samples were collected and each water sample was analyzed for U, and each sediment sample was analyzed for 43 elements, including U and Th. Uranium concentrations in water samples range from below the detection limit of 0.02 ppB to 702.26 ppB and have a median of 1.73 ppB and a mean of 11.76 ppB. Water samples containing high uranium concentrations generally are associated with known uranium mining activity or units known to be uranium bearing. About one-third of the water samples containing high uranium concentrations were collected from locations within the Pumpkin Buttes and Turnercrest-Ross Districts. Nearlymore » half of the water samples containing high uranium concentrations were collected from locations just west of the Monument Hill and Highland Flats-Box Creek Districts. Similar anomalous uranium concentrations in this region have been reported updip from Exxon's Highland uranium deposits. High uranium concentrations were also found associated with the Lance Creek-Old Woman Anticline District. Uranium concentrations in sediment samples range from 1.14 to 220.70 ppM and have a median of 3.37 ppM and a mean of 4.03 ppM. Throughout the major uranium mining districts of the Powder River Basin, sediment samples with high uranium concentrations were collected from dry streams located near wells producing water samples with high uranium concentrations. High uranium concentrations were also found associated with the Lance Creek oil field where uranium mineralization is known in the White River formation. High uranium concentrations were also found in sediment samples in areas where uranium mineralization is not known. These samples are from dry streams in areas underlain by the White River formation, the Niobrara formation, and the Pierre, Carlisle, Belle Fourche, and Mowry shales.« less

Azimuthal anisotropy in U+U collisions at STAR

DOE PAGES

Wang, Hui; Sorensen, Paul

2014-10-06

The azimuthal anisotropy of particle production is commonly used in high-energy nuclear collisions to study the early evolution of the expanding system. The prolate shape of uranium nuclei makes it possible to study how the geometry of the colliding nuclei affects final state anisotropies. It also provides a unique opportunity to understand how entropy is produced in heavy ion collisions. In this paper, the two- and four- particle cumulant v 2 (v 2{2} and v 2{4}) from U+U collisions at √ sNN = 193 GeV and Au+Au collisions at √ sNN = 200 GeV for inclusive charged hadrons will bemore » presented. The STAR Zero Degree Calorimeters are used to select very central collisions. Differences were observed between the multiplicity dependence of v 2{2} for most central Au+Au and U+U collisions. The multiplicity dependence of v 2{2} in central collisions were compared to Monte Carlo Glauber model predictions and it was seen that this model cannot explain the present results. (auth)« less

Materials and Methods for Streamlined Laboratory Analysis of Environmental Samples, FY 2016 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Addleman, Raymond S.; Naes, Benjamin E.; McNamara, Bruce K.

The International Atomic Energy Agency (IAEA) relies upon laboratory analysis of environmental samples (typically referred to as “swipes”) collected during on-site inspections of safeguarded facilities to support the detection and deterrence of undeclared activities. Unfortunately, chemical processing and assay of the samples is slow and expensive. A rapid, effective, and simple extraction process and analysis method is needed to provide certified results with improved timeliness at reduced costs (principally in the form of reduced labor), while maintaining or improving sensitivity and efficacy. To address these safeguard needs the Pacific Northwest National Laboratory (PNNL) explored and demonstrated improved methods for environmentalmore » sample (ES) analysis. Improvements for both bulk and particle analysis were explored. To facilitate continuity and adoption, the new sampling materials and processing methods will be compatible with existing IAEA protocols for ES analysis. PNNL collaborated with Oak Ridge National Laboratory (ORNL), which performed independent validation of the new bulk analysis methods and compared performance to traditional IAEA’s Network of Analytical Laboratories (NWAL) protocol. ORNL efforts are reported separately. This report describes PNNL’s FY 2016 progress, which was focused on analytical application supporting environmental monitoring of uranium enrichment plants and nuclear fuel processing. In the future the technology could be applied to other safeguard applications and analytes related to fuel manufacturing, reprocessing, etc. PNNL’s FY 2016 efforts were broken into two tasks and a summary of progress, accomplishments and highlights are provided below. Principal progress and accomplishments on Task 1, Optimize Materials and Methods for ICP-MS Environmental Sample Analysis, are listed below. • Completed initial procedure for rapid uranium extraction from ES swipes based upon carbonate-peroxide chemistry (delivered to ORNL for evaluation). • Explored improvements to carbonate-peroxide rapid uranium extraction chemistry. • Evaluated new sampling materials and methods (in collaboration with ORNL). • Demonstrated successful ES extractions from standard and novel swipes for a wide range uranium compounds of interest including UO 2F 2 and UO 2(NO 3) 2, U 3O 8 and uranium ore concentrate. • Completed initial discussions with commercial suppliers of PTFE swipe materials. • Submitted one manuscript for publication. Two additional drafts are being prepared. Principal progress and accomplishments on Task 2, Optimize Materials and Methods for Direct SIMS Environmental Sample Analysis, are listed below. • Designed a SIMS swipe sample holder that retrofits into existing equipment and provides simple, effective, and rapid mounting of ES samples for direct assay while enabling automation and laboratory integration. • Identified preferred conductive sampling materials with better performance characteristics. • Ran samples on the new PNNL NWAL equivalent Cameca 1280 SIMS system. • Obtained excellent agreement between isotopic ratios for certified materials and direct SIMS assay of very low levels of LEU and HEU UO 2F 2 particles on carbon fiber sampling material. Sample activities range from 1 to 500 CPM (uranium mass on sample is dependent upon specific isotope ratio but is frequently in the subnanogram range). • Found that the presence of the UF molecular ions, as measured by SIMS, provides chemical information about the particle that is separate from the uranium isotopics and strongly suggests that those particles originated from an UF6 enrichment activity. • Submitted one manuscript for publication. Another manuscript is in preparation.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE PAGES

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-06-13

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

The oxidative corrosion of carbide inclusions at the surface of uranium metal during exposure to water vapour.

PubMed

Scott, T B; Petherbridge, J R; Harker, N J; Ball, R J; Heard, P J; Glascott, J; Allen, G C

2011-11-15

The reaction between uranium and water vapour has been well investigated, however discrepancies exist between the described kinetic laws, pressure dependence of the reaction rate constant and activation energies. Here this problem is looked at by examining the influence of impurities in the form of carbide inclusions on the reaction. Samples of uranium containing 600 ppm carbon were analysed during and after exposure to water vapour at 19 mbar pressure, in an environmental scanning electron microscope (ESEM) system. After water exposure, samples were analysed using secondary ion mass spectrometry (SIMS), focused ion beam (FIB) imaging and sectioning and transmission electron microscopy (TEM) with X-ray diffraction (micro-XRD). The results of the current study indicate that carbide particles on the surface of uranium readily react with water vapour to form voluminous UO(3) · xH(2)O growths at rates significantly faster than that of the metal. The observation may also have implications for previous experimental studies of uranium-water interactions, where the presence of differing levels of undetected carbide may partly account for the discrepancies observed between datasets. Crown Copyright © 2011. Published by Elsevier B.V. All rights reserved.

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Johnson, J. A.; Collins, J. L.; McMurray, J. W.; Reif, T. J.; Brown, D. R.

2018-01-01

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC2), which is UC1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UC2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90-92% of TD with full conversion of UC to UC2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC2. The selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

DOE PAGES

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee; ...

2017-10-12

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Method and system for improved resolution of a compensated calorimeter detector

DOEpatents

Dawson, John W.

1991-01-01

An improved method and system for a depleted uranium calorimeter detector used in high energy physics experiments. In a depleted uranium calorimeter detector, the energy of a particle entering the calorimeter detector is determined and the output response of the calorimeter detector is compensated so that the ratio of the integrated response of the calorimeter detector from a lepton to the integrated response of the calorimeter detector from a hadron of the same energy as the lepton is approximately equal to 1. In the present invention, the energy of a particle entering the calorimeter detector is determined as a function of time and the hadron content of the response of the calorimeter detector is inferred based upon the time structure of the energy pulse measured by the calorimeter detector. The energy measurement can be corrected based on the inference of the hadron content whereby the resolution of the calorimeter can be improved.

Depleted and enriched uranium exposure quantified in former factory workers and local residents of NL Industries, Colonie, NY USA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnason, John G.

Background: Between 1958 and 1982, NL Industries manufactured components of enriched (EU) and depleted uranium (DU) at a factory in Colonie NY, USA. More than 5 metric tons of DU was deposited as microscopic DU oxide particles on the plant site and surrounding residential community. A prior study involving a small number of individuals (n=23) indicated some residents were exposed to DU and former workers to both DU and EU, most probably through inhalation of aerosol particles. Objectives: Our aim was to measure total uranium [U] and the uranium isotope ratios: {sup 234}U/{sup 238}U; {sup 235}U/{sup 238}U; and {sup 236}U/{supmore » 238}U, in the urine of a cohort of former workers and nearby residents of the NLI factory, to characterize individual exposure to natural uranium (NU), DU, and EU more than 3 decades after production ceased. Methods: We conducted a biomonitoring study in a larger cohort of 32 former workers and 99 residents, who may have been exposed during its period of operation, by measuring Total U, NU, DU, and EU in urine using Sector Field Inductively Coupled Plasma - Mass Spectrometry (SF-ICP-MS). Results: Among workers, 84% were exposed to DU, 9% to EU and DU, and 6% to natural uranium (NU) only. For those exposed to DU, urinary isotopic and [U] compositions result from binary mixing of NU and the DU plant feedstock. Among residents, 8% show evidence of DU exposure, whereas none shows evidence of EU exposure. For residents, the [U] geometric mean is significantly below the value reported for NHANES. There is no significant difference in [U] between exposed and unexposed residents, suggesting that [U] alone is not a reliable indicator of exposure to DU in this group. Conclusions: Ninety four percent of workers tested showed evidence of exposure to DU, EU or both, and were still excreting DU and EU decades after leaving the workforce. The study demonstrates the advantage of measuring multiple isotopic ratios (e.g., {sup 236}U/{sup 238}U and {sup 235}U/{sup 238}U) over a single ratio ({sup 235}U/{sup 238}U) in determining sources of uranium exposure. - Highlights: • Biomonitoring study of residents and former workers exposed to DU in Colonie NY. • Urine (99 residents+32 former workers) analyzed for depleted uranium (DU). • DU detected in 84% of workers and 8% of residents >30 years after plant closed. • Enriched uranium detected in 6% of former workers based on isotope ratios.« less

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabin, S.A.; Lotts, A.L.; Hammond, J.P.

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodyingmore » hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)« less

Irradiation performance of AGR-1 high temperature reactor fuel

DOE PAGES

Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; ...

2015-10-23

The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that itmore » was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10 –4 to 5 × 10 –4 for 154Eu and 8 × 10 –7 to 3 × 10 –5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10 –6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10 5 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10 –5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization of these elements within the SiC microstructure is the subject of ongoing focused study.« less

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Code of Federal Regulations, 2010 CFR

2010-01-01

... and special nuclear material in the accounting records are based on measured values; (3) A measurement... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for uranium... Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

10 CFR 20.2206 - Reports of individual monitoring.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Energy NUCLEAR REGULATORY COMMISSION STANDARDS FOR PROTECTION AGAINST RADIATION Reports § 20.2206 Reports...) Operate a nuclear reactor designed to produce electrical or heat energy pursuant to § 50.21(b) or § 50.22... nuclear material in a quantity exceeding 5,000 grams of contained uranium-235, uranium-233, or plutonium...

10 CFR 20.2206 - Reports of individual monitoring.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Energy NUCLEAR REGULATORY COMMISSION STANDARDS FOR PROTECTION AGAINST RADIATION Reports § 20.2206 Reports...) Operate a nuclear reactor designed to produce electrical or heat energy pursuant to § 50.21(b) or § 50.22... nuclear material in a quantity exceeding 5,000 grams of contained uranium-235, uranium-233, or plutonium...

10 CFR 20.2206 - Reports of individual monitoring.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Energy NUCLEAR REGULATORY COMMISSION STANDARDS FOR PROTECTION AGAINST RADIATION Reports § 20.2206 Reports...) Operate a nuclear reactor designed to produce electrical or heat energy pursuant to § 50.21(b) or § 50.22... nuclear material in a quantity exceeding 5,000 grams of contained uranium-235, uranium-233, or plutonium...

40 CFR 440.34 - New source performance standards (NSPS).

Code of Federal Regulations, 2013 CFR

2013-07-01

... underground, that produce uranium ore, excluding mines using in situ leach methods, shall not exceed: Effluent... leach process for the extraction of uranium or from mines and mills using in situ leach methods. The... Vanadium Ores Subcategory § 440.34 New source performance standards (NSPS). Except as provided in subpart L...

40 CFR 440.34 - New source performance standards (NSPS).

Code of Federal Regulations, 2012 CFR

2012-07-01

... underground, that produce uranium ore, excluding mines using in situ leach methods, shall not exceed: Effluent... leach process for the extraction of uranium or from mines and mills using in situ leach methods. The... Vanadium Ores Subcategory § 440.34 New source performance standards (NSPS). Except as provided in subpart L...

40 CFR 440.34 - New source performance standards (NSPS).

Code of Federal Regulations, 2014 CFR

2014-07-01

... underground, that produce uranium ore, excluding mines using in situ leach methods, shall not exceed: Effluent... leach process for the extraction of uranium or from mines and mills using in situ leach methods. The... Vanadium Ores Subcategory § 440.34 New source performance standards (NSPS). Except as provided in subpart L...

Electron-spectroscopy studies of clean thorium and uranium surfaces. Chemisorption and initial stages of reaction with O2, CO, and CO2

NASA Astrophysics Data System (ADS)

McLean, W.; Colmenares, C. A.; Smith, R. L.; Somorjai, G. A.

1982-01-01

The adsorption of O2, CO, and CO2 on the thorium (111) crystal face and on polycrystalline α-uranium has been investigated by x-ray photoelectron spectroscopy, Auger electron spectroscopy (AES), and secondary-ion mass spectroscopy (SIMS) at 300 K. Oxygen adsorption on both metals resulted in the formation of the metal dioxide. CO and CO2 adsorption on Th(111) produced species derived from atomic carbon and oxygen; the presence of molecular CO was also detected. Only atomic carbon and oxygen were observed on uranium. Elemental depth profiles by AES and SIMS indicated that the carbon produced by the dissociation of CO or CO2 diffused into the bulk of the metals to form a carbide, while the oxygen remained on their surfaces as an oxide.

Efficacy of Biostimulation for Uranium Sequestration: Coupled Effects Sediment/Groundwater Geochemistry and Microbiology

NASA Astrophysics Data System (ADS)

Xu, J.; Veeramani, H.; Qafoku, N. P.; Singh, G.; Pruden, A.; Kukkadapu, R. K.; Hochella, M. F., Jr.

2015-12-01

A systematic flow-through column study was conducted using sediments and groundwater from the subsurface at the U.S. Department of Energy's Integrated Field Research Challenge (IFRC) site in Rifle, Colorado, to better understand the efficacy of uranium removal from the groundwater with and without biostimulation in the form of acetate amendments. The interactive effects of acetate amendment, groundwater/sediment geochemistry, and intrinsic bacterial community composition were evaluated using four types of sediments, collected from different uranium-contaminated (D08, LQ107, CD) or non-contaminated (RABS) aquifers. Subtle variations in the sediments' geochemistry in terms of mineral compositions, particle sizes, redox conditions, and metal(loid) co-contaminants had a marked effect on the uranium removal efficiency, following a descending trend of D08 (~ 90 to 95%) >> RABS (~ 20 to 25) ≥ LQ107 (~ 15 to 20%) > CD (~ -10 to 0%). Overall, biostimulation of the sediments with acetate drove deeper anoxic conditions and observable shifts in bacterial population structures. The abundance of dissimilatory sulfate-reduction genes (i.e., drsA), markers of sulfate-reducing bacteria, were highest in the sediments that performed best in terms of uranium removal. By comparison, no obvious associations were found between the uranium removal efficiency and the abundance of typical iron-reducing microorganisms, e.g., Geobacter spp. In the sediments where bacterial biomass was relatively low and sulfate-reduction was not detected (i.e., CD), abiotic adsorption onto fine mineral surfaces such as phyllosilates likely played a dominant role in the attenuation of aqueous uranium. In these scenarios, however, acetate amendment induced significant remobilization of the sequestered uranium and other heavy metals (e.g., strontium), leading to zero or negative uranium removal efficiencies (i.e., CD). The results of this study suggest that reductive immobilization of uranium can be effectively achieved under predominantly sulfate-reducing conditions in sediment microenvironments when bioavailable iron (III) (oxyhydr)oxides are mostly depleted, and provide insight into the integrated roles of sediment geochemistry, mineralogy, and bacterial population dynamics.

Infrared spectroscopy of extreme coordination: the carbonyls of U(+) and UO(2)(+).

PubMed

Ricks, Allen M; Gagliardi, Laura; Duncan, Michael A

2010-11-17

Uranium and uranium dioxide carbonyl cations produced by laser vaporization are studied with mass-selected ion infrared spectroscopy in the C-O stretching region. Dissociation patterns, spectra, and quantum chemical calculations establish that the fully coordinated ions are U(CO)(8)(+) and UO(2)(CO)(5)(+), with D(4d) square antiprism and D(5h) pentagonal bipyramid structures. Back-bonding in U(CO)(8)(+) causes a red-shifted CO stretch, but back-donation is inefficient for UO(2)(CO)(5)(+), producing a blue-shifted CO stretch characteristic of nonclassical carbonyls.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour.

PubMed

Martin, T L; Coe, C; Bagot, P A J; Morrall, P; Smith, G D W; Scott, T; Moody, M P

2016-07-12

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

NASA Astrophysics Data System (ADS)

Martin, T. L.; Coe, C.; Bagot, P. A. J.; Morrall, P.; Smith, G. D. W.; Scott, T.; Moody, M. P.

2016-07-01

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

Identifying anthropogenic uranium compounds using soft X-ray near-edge absorption spectroscopy

NASA Astrophysics Data System (ADS)

Ward, Jesse D.; Bowden, Mark; Tom Resch, C.; Eiden, Gregory C.; Pemmaraju, C. D.; Prendergast, David; Duffin, Andrew M.

2017-01-01

Uranium ores mined for industrial use are typically acid-leached to produce yellowcake and then converted into uranium halides for enrichment and purification. These anthropogenic chemical forms of uranium are distinct from their mineral counterparts. The purpose of this study is to use soft X-ray absorption spectroscopy to characterize several common anthropogenic uranium compounds important to the nuclear fuel cycle. Chemical analyses of these compounds are important for process and environmental monitoring. X-ray absorption techniques have several advantages in this regard, including element-specificity, chemical sensitivity, and high spectral resolution. Oxygen K-edge spectra were collected for uranyl nitrate, uranyl fluoride, and uranyl chloride, and fluorine K-edge spectra were collected for uranyl fluoride and uranium tetrafluoride. Interpretation of the data is aided by comparisons to calculated spectra. The effect of hydration state on the sample, a potential complication in interpreting oxygen K-edge spectra, is discussed. These compounds have unique spectral signatures that can be used to identify unknown samples.

The uranium deposit at the Yellow Canary claims, Daggett County, Utah

USGS Publications Warehouse

Wilmarth, V.R.; Vickers, R.C.; McKeown, F.A.; Beroni, E.P.

1952-01-01

The Yellow Canary claims uranium deposit is on the west side of Red Creek Canyon in the northern part of the Uinta Mountains, Daggett County, Utah. The claims have been developed by two adits, three open cuts, and several hundred deep of bulldozer trenches. No uranium ore has been produced from this deposit. The uranium deposit at the Yellow Canary claims is in the Red Creek quartzite of pre-Cambrian age. The formation is composed of intercalated beds of quartzite, hornblendite, garnet schist, staurolite schist, and quartz-mica schist and is intruded by diorite dikes. A thick unit of highly fractured white quatrzite at the top of the formation contains tyutamunite as coatings on fracture surfaces. The tyutamunite is associated with carnotite, volborthite, iron oxides, azurite, malachite, brochantite, and hyalite. The secondary uranium and vanadium minerals are believed to be alteration products of primary minerals. The uranium content of 15 samples from this property ranged from 0.000 to 0.57 percent.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

PubMed Central

Martin, T. L.; Coe, C.; Bagot, P. A. J.; Morrall, P.; Smith, G. D. W; Scott, T.; Moody, M. P.

2016-01-01

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour. PMID:27403638

Alpha-particle spectrometer experiment

NASA Technical Reports Server (NTRS)

Gorenstein, P.; Bjorkholm, P.

1972-01-01

Mapping the radon emanation of the moon was studied to find potential areas of high activity by detection of radon isotopes and their daughter products. It was felt that based on observation of regions overflown by Apollo spacecraft and within the field of view of the alpha-particle spectrometer, a radon map could be constructed, identifying and locating lunar areas of outgassing. The basic theory of radon migration from natural concentrations of uranium and thorium is discussed in terms of radon decay and the production of alpha particles. The preliminary analysis of the results indicates no significant alpha emission.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

PRODUCTION OF URANIUM TUBING

DOEpatents

Creutz, E.C.

1958-04-15

The manufacture of thin-walled uranium tubing by the hot-piercing techique is described. Uranium billets are preheated to a temperature above 780 d C. The heated billet is fed to a station where it is engaged on its external surface by three convex-surfaced rotating rollers which are set at an angle to the axis of the billet to produce a surface friction force in one direction to force the billet over a piercing mandrel. While being formed around the mandrel and before losing the desired shape, the tube thus formed is cooled by a water spray.

Uraniferous opal, Virgin Valley, Nevada: conditions of formation and implications for uranium exploration

USGS Publications Warehouse

Zielinski, R.A.

1982-01-01

Uraniferous, fluorescent opal, which occurs in tuffaceous sedimentary rocks at Virgin Valley, Nevada, records the temperature and composition of uranium-rich solutions as well as the time of uranium-silica coprecipitation. Results are integrated with previous geologic and geochronologic data for the area to produce a model for uranium mobility that may be used to explore for uranium deposits in similar geologic settings. Uraniferous opal occurs as replacements of diatomite, or silicic air-fall ash layers in tuffaceous lakebeds of the Virgin Valley Formation (Miocene) of Merriam (1907). Fission-track radiography shows uranium to be homogeneously dispersed throughout the opal structure, suggesting coprecipitation of dissolved uranium and silica gel. Fluid inclusions preserved within opal replacements of diatomite have homogenization temperatures in the epithermal range and are of low salinity. Four samples of opal from one locality all have U-Pb apparent ages which suggest uraniferous opal precipitation in late Pliocene time. These ages correspond to a period of local, normal faulting, and highangle faults may have served as vertical conduits for transport of deep, thermalized ground water to shallower levels. Lateral migration of rising solutions occurred at intersections of faults with permeable strata. Silica and some uranium were dissolved from silica-rich host strata of 5-20 ppm original uranium content and reprecipitated as the solutions cooled. The model predicts that in similar geologic settings, ore-grade concentrations of uranium will occur in permeable strata that intersect high-angle faults and that contain uranium source rocks as well as efficient reductant traps for uranium. In the absence of sufficient quantities of reductant materials, uranium will be flushed from the system or will accumulate in low-grade disseminated hosts such as uraniferous opal. ?? 1982.

Neutron detector using sol-gel absorber

DOEpatents

Hiller, John M.; Wallace, Steven A.; Dai, Sheng

1999-01-01

An neutron detector composed of fissionable material having ions of lithium, uranium, thorium, plutonium, or neptunium, contained within a glass film fabricated using a sol-gel method combined with a particle detector is disclosed. When the glass film is bombarded with neutrons, the fissionable material emits fission particles and electrons. Prompt emitting activated elements yielding a high energy electron contained within a sol-gel glass film in combination with a particle detector is also disclosed. The emissions resulting from neutron bombardment can then be detected using standard UV and particle detection methods well known in the art, such as microchannel plates, channeltrons, and silicon avalanche photodiodes.

Yellow Canary uranium deposits, Daggett County, Utah

USGS Publications Warehouse

Wilmarth, Verl Richard

1953-01-01

The Yellow Canary uranium deposit is on the west side of Red Creek Canyon in the northern part of the Uinta Mountains, Daggett County, Utah. Two claims have been developed by means of an adit, three opencuts, and several hundred feet of bulldozer trenches. No uranium ore has been produced from this deposit. The deposit is in the pre-Cambrian Red Creek quartzite. This formation is composed of intercalated beds of quartzite, hornblendite, garnet schist, staurolite schist, and quartz-mica schist and is intruded by dioritic dikes. A thick unit of highly fractured white quartzite near the top of the formation contains tyuyamunite as coatings on fracture surfaces. The tyuyamunite is associated with carnotite, volborthite, iron oxides, azurite, malachite, brochantite, and hyalite. The uranium and vanadium minerals are probably alteration products of primary minerals. The uranium content of 15 samples from this property ranged from 0.000 to 0.57 percent.

Neutron-rich isotope production using a uranium carbide - carbon nanotubes SPES target prototype

NASA Astrophysics Data System (ADS)

Corradetti, S.; Biasetto, L.; Manzolaro, M.; Scarpa, D.; Carturan, S.; Andrighetto, A.; Prete, G.; Vasquez, J.; Zanonato, P.; Colombo, P.; Jost, C. U.; Stracener, D. W.

2013-05-01

The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

The estuarine chemistry and isotope systematics of 234,238U in the Amazon and Fly Rivers

USGS Publications Warehouse

Swarzenski, P.; Campbell, P.; Porcelli, D.; McKee, B.

2004-01-01

Natural concentrations of 238U and ??234U values were determined in estuarine surface waters and pore waters of the Amazon and Fly (Papua New Guinea) Rivers to investigate U transport phenomena across river-dominated land-sea margins. Discharge from large, tropical rivers is a major source of dissolved and solid materials transported to the oceans, and are important in defining not only oceanic mass budgets, but also terrestrial weathering rates. On the Amazon shelf, salinity-property plots of dissolved organic carbon, pH and total suspended matter revealed two vastly contrasting water masses that were energetically mixed. In this mixing zone, the distribution of uranium was highly non-conservative and exhibited extensive removal from the water column. Uranium removal was most pronounced within a salinity range of 0-16.6, and likely the result of scavenging and flocculation reactions with inorganic (i.e., Fe/Mn oxides) and organic colloids/particles. Removal of uranium may also be closely coupled to exchange and resuspension processes at the sediment/water interface. An inner-shelf pore water profile indicated the following diagenetic processes: extensive (???1 m) zones of Fe(III) - and, to a lesser degree, Mn(IV) - reduction in the absence of significant S(II) concentrations appeared to facilitate the formation of various authigenic minerals (e.g., siderite, rhodocrosite and uraninite). The pore water dissolved 238U profile co-varied closely with Mn(II). Isotopic variations as evidenced in ??234U pore waters values from this site revealed information on the origin and history of particulate uranium. Only after a depth of about 1 m did the ??234U value approach unity (secular equilibrium), denoting a residual lattice bound uranium complex that is likely an upper-drainage basin weathering product. This suggests that the enriched ??234U values represent a riverine surface complexation product that is actively involved in Mn-Fe diagenetic cycles and surface complexation reactions. In the Fly River estuary, 238U appears to exhibit a reasonably conservative distribution as a function of salinity. The absence of observed U removal does not necessarily imply non-reactivity, but instead may record an integration of concurrent U removal and release processes. There is not a linear correlation between ??234U vs. 1/ 238U that would imply simple two component mixing. It is likely that resuspension of bottom sediments, prolonged residence times in the lower reaches of the Fly River, and energetic particle-colloid interactions contribute to the observed estuarine U distribution. The supply of uranium discharged from humid, tropical river systems to the sea appears to be foremost influenced by particle/water interactions that are ultimately governed by the particular physiographic and hydrologic characteristics of an estuary. ?? 2004 Elsevier Ltd. All rights reserved.

78 FR 21100 - Low Enriched Uranium From France: Final Results of the Expedited Second Sunset Review of the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-09

... received no response from the respondent interested parties, i.e., French uranium producers and exporters... Centralized Electronic Service System (IA ACCESS). IA ACCESS is available to registered users at http... the Internet at http://trade.gov/ia/ . The signed Decision Memorandum and electronic versions of the...

Depleted UF6 Internet Resources

Science.gov Websites

been used to color glass for almost 2 millennia. A uranium-colored glass object was found near Naples , Italy, and dated to about 79 A.D. Uranium oxide added to glass produces a yellow to greenish hue. more Board Defense Nuclear Facilities Safety Board (DNFSB) The Defense Nuclear Facilities Safety Board

Lung dosimetry for inhaled long-lived radionuclides and radon progeny.

PubMed

Hussain, M; Winkler-Heil, R; Hofmann, W

2011-05-01

The current version of the stochastic lung dosimetry model IDEAL-DOSE considers deposition in the whole tracheobronchial (TB) and alveolar airway system, while clearance is restricted to TB airways. For the investigation of doses produced by inhaled long-lived radionuclides (LLR) together with short-lived radon progeny, alveolar clearance has to be considered. Thus, present dose calculations are based on the average transport rates proposed for the revision of the ICRP human respiratory tract model. The results obtained indicate that LLR cleared from the alveolar region can deliver up to two to six times higher doses to the TB region when compared with the doses from directly deposited particles. Comparison of LLR doses with those of short-lived radon progeny indicates that LLR in uranium mines can deliver up to 5 % of the doses predicted for the short-lived radon daughters.

Spacecraft-produced neutron fluxes on Skylab

NASA Technical Reports Server (NTRS)

Quist, T. C.; Furst, M.; Burnett, D. S.; Baum, J. H.; Peacock, C. L., Jr.; Perry, D. G.

1977-01-01

Estimates of neutron fluxes in different energy ranges are reported for the Skylab spacecraft. Detectors composed of uranium, thorium, and bismuth foils with mica as a fission track recorder, as well as boron foils with cellulose acetate as an alpha-particle recorder, were deployed at different positions in the Orbital Workshop. It was found that the Skylab neutron flux was dominated by high energy (greater than 1 MeV) contributions and that there was no significant time variation in the fluxes. Firm upper limits of 7-15 neutrons/sq cm-sec, depending on the detector location in the spacecraft, were established for fluxes above 1 MeV. Below 1 MeV, the neutron fluxes were about an order of magnitude lower. The neutrons are interpreted as originating from the interactions of leakage protons from the radiation belt with the spacecraft.

Fuel preparation for use in the production of medical isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

Evolution of uranium distribution and speciation in mill tailings, COMINAK Mine, Niger.

PubMed

Déjeant, Adrien; Galoisy, Laurence; Roy, Régis; Calas, Georges; Boekhout, Flora; Phrommavanh, Vannapha; Descostes, Michael

2016-03-01

This study investigated the evolution of uranium distribution and speciation in mill tailings from the COMINAK mine (Niger), in production since 1978. A multi-scale approach was used, which combined high resolution remote sensing imagery, ICP-MS bulk rock analyses, powder X-ray diffraction, Scanning Electron Microscopy, Focused Ion Beam--Transmission Electron Microscopy and X-ray Absorption Near Edge Spectroscopy. Mineralogical analyses showed that some ore minerals, including residual uraninite and coffinite, undergo alteration and dissolution during tailings storage. The migration of uranium and other contaminants depends on (i) the chemical stability of secondary phases and sorbed species (dissolution and desorption processes), and (ii) the mechanical transport of fine particles bearing these elements. Uranium is stabilized after formation of secondary uranyl sulfates and phosphates, and adsorbed complexes on mineral surfaces (e.g. clay minerals). In particular, the stock of insoluble uranyl phosphates increases with time, thus contributing to the long-term stabilization of uranium. At the surface, a sulfate-cemented duricrust is formed after evaporation of pore water. This duricrust limits water infiltration and dust aerial dispersion, though it is enriched in uranium and many other elements, because of pore water rising from underlying levels by capillary action. Satellite images provided a detailed description of the tailings pile over time and allow monitoring of the chronology of successive tailings deposits. Satellite images suggest that uranium anomalies that occur at deep levels in the pile are most likely former surface duricrusts that have been buried under more recent tailings. Copyright © 2015 Elsevier B.V. All rights reserved.

Structure and Reactivity of X-ray Amorphous Uranyl Peroxide, U 2O 7

DOE PAGES

Odoh, Samuel O.; Shamblin, Jacob; Colla, Christopher A.; ...

2016-03-14

Recent accidents resulting in worker injury and radioactive contamination occurred due to pressurization of uranium yellowcake drums produced in the western USA. The drums contained an unexpected X-ray amorphous reactive form of uranium oxide, U 2O7. Heating hydrated uranyl peroxides produced during in situ mining unintentionally produced U 2O 7. It is a hygroscopic anhydrous uranyl peroxide that reacts rapidly with water to release O 2 gas and form metaschoepite, a uranyl-oxide hydrate. Quantum chemical calculations indicate that the most stable U 2O 7 conformer consists of two bent (UO 2) 2+ uranyl ions bridged by a peroxide group bidentatemore » and parallel to each uranyl ion, and a μ2-O atom, resulting in charge neutrality. A pair distribution function from neutron total scattering supports this structural model. The reactivity of U 2O 7 in water and with water in air is much higher than other uranium oxides, and this can be both hazardous and potentially advantageous in the nuclear fuel cycle.« less

In-line assay monitor for uranium hexafluoride

DOEpatents

Wallace, Steven A.

1981-01-01

An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The monitor is intended for uses such as safeguard applications to assure that weapons grade uranium is not being produced in an enrichment cascade. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from the uranium-235 present in the specimen. Simultaneously, the gamma emissions from the uranium-235 of the specimen and the source emissions transmitted through the sample are counted and stored in a multiple channel analyzer. The uranium-235 content of the specimen is determined from the comparison of the accumulated 185 keV energy counts and the reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen. The process eliminates the necessity of knowing the system operating conditions and yet obtains the necessary data without need for large scintillation crystals and sophisticated mechanical designs.

Innovative Elution Processes for Recovering Uranium and Transition Metals from Amidoxime-based Adsorbents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wai, Chien M.

Amidoxime-based polymer fibers are considered one of the most promising materials for sequestering uranium from seawater. The high-surface-area polymer fibers containing amidoxime and carboxylate groups synthesized by Oak Ridge National Lab (ORNL-AF1) show very high uranium adsorption capacities known in the literature. Effective elution of uranium and repeated use of the adsorbent are important factors affecting the cost of producing uranium from seawater using this material. Traditional acid leaching of uranium followed by KOH conditioning of the fiber causes chemical changes and physical damage to the ORNL-AF1 adsorbent. Two alkaline solution leaching methods were developed by this project, one usesmore » a highly concentrated (3 M) potassium bicarbonate solution at pH 8.3 and 40 °C; the other uses a mixture of sodium carbonate and hydrogen peroxide at pH 10.4. Both elution methods do not require KOH conditioning prior to reusing the fiber adsorbent. The conditions of eluting uranium from the amidoxime-based adsorbent using these alkaline solutions are confirmed by thermodynamic calculations. The bicarbonate elution method is selective for uranium recovery compared to other elution methods and causes no chemical change to the fiber material based on FTIR spectroscopy« less

Uranium Mining and Norm in North America-Some Perspectives on Occupational Radiation Exposure.

PubMed

Brown, Steven H; Chambers, Douglas B

2017-07-01

All soils and rocks contain naturally occurring radioactive materials (NORM). Many ores and raw materials contain relatively elevated levels of natural radionuclides, and processing such materials can further increase the concentrations of naturally occurring radionuclides. In the U.S., these materials are sometimes referred to as technologically-enhanced naturally occurring radioactive materials (TENORM). Examples of NORM minerals include uranium ores, monazite (a source of rare earth minerals), and phosphate rock used to produce phosphate fertilizer. The processing of these materials has the potential to result in above-background radiation exposure to workers. Following a brief review of the sources and potential for worker exposure from NORM in these varied industries, this paper will then present an overview of uranium mining and recovery in North America, including discussion on the mining methods currently being used for both conventional (underground, open pit) and in situ leach (ISL), also referred to as In Situ Recovery (ISR), and the production of NORM materials and wastes associated with these uranium recovery methods. The radiological composition of the NORM products and wastes produced and recent data on radiological exposures received by workers in the North American uranium recovery industry are then described. The paper also identifies the responsible government agencies in the U.S. and Canada assigned the authority to regulate and control occupational exposure from these NORM materials.

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility.

PubMed

Döppner, T; Callahan, D A; Hurricane, O A; Hinkel, D E; Ma, T; Park, H-S; Berzak Hopkins, L F; Casey, D T; Celliers, P; Dewald, E L; Dittrich, T R; Haan, S W; Kritcher, A L; MacPhee, A; Le Pape, S; Pak, A; Patel, P K; Springer, P T; Salmonson, J D; Tommasini, R; Benedetti, L R; Bond, E; Bradley, D K; Caggiano, J; Church, J; Dixit, S; Edgell, D; Edwards, M J; Fittinghoff, D N; Frenje, J; Gatu Johnson, M; Grim, G; Hatarik, R; Havre, M; Herrmann, H; Izumi, N; Khan, S F; Kline, J L; Knauer, J; Kyrala, G A; Landen, O L; Merrill, F E; Moody, J; Moore, A S; Nikroo, A; Ralph, J E; Remington, B A; Robey, H F; Sayre, D; Schneider, M; Streckert, H; Town, R; Turnbull, D; Volegov, P L; Wan, A; Widmann, K; Wilde, C H; Yeamans, C

2015-07-31

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a "high-foot" laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shape closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10^{16} neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.

Leaching behavior of U, Mn, Sr, and Pb from different particle-size fractions of uranium mill tailings.

PubMed

Liu, Bo; Peng, Tongjiang; Sun, Hongjuan

2017-06-01

Pollution by the release of heavy metals from tailings constitutes a potential threat to the environment. To characterize the processes governing the release of Mn, Sr, Pb, and U from the uranium mill tailings, a dynamic leaching test was applied for different size of uranium mill tailings samples. Inductively coupled plasma atomic emission spectroscopy (ICP-AES) and inductively coupled plasma mass spectrometry (ICP-MS) were performed to determine the content of Mn, Sr, Pb, and U in the leachates. The release of mobile Mn, Sr, Pb, and U fraction was slow, being faster in the initial stage and then attained a near steady-state condition. The experimental results demonstrate that the release of Mn, Sr, Pb, and U from uranium mill tailings with different size fractions is controlled by a variety of mechanisms. Surface wash-off is the release mechanism for Mn. The main release mechanism of Sr and Pb is the dissolution in the initial leaching stage. For U, a mixed process of wash-off and diffusion is the controlling mechanism.

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Döppner, T.; Callahan, D. A.; Hurricane, O. A.

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore » closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10 16 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

DOE PAGES

Döppner, T.; Callahan, D. A.; Hurricane, O. A.; ...

2015-07-28

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore » closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10 16 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

Measurements of n-p correlations in the reaction of relativistic neon with uranium

NASA Technical Reports Server (NTRS)

Frankel, K.; Schimmerling, W.; Rasmussen, J. O.; Crowe, K. M.; Bistirlich, J.; Bowman, H.; Hashimoto, O.; Murphy, D. L.; Ridout, J.; Sullivan, J. P.;

Radiation protection design for the Super-FRS and SIS100 at the international FAIR facility

NASA Astrophysics Data System (ADS)

Kozlova, Ekaterina; Sokolov, Alexey; Radon, Torsten; Lang, Rupert; Conrad, Inna; Fehrenbacher, Georg; Weick, Helmut; Winkler, Martin

2017-09-01

The new accelerator SIS100 and the Super-FRS will be built at the international Facility for Antiprotons and Ion Research FAIR. The synchrotron SIS100 is a core part of the FAIR facility which serves for acceleration of ions like Uranium up to 2.7 GeV/u with intensities of 3x1011 particles per second or protons up to 30 GeV with intensities of 5x1012 particles per second. The Super-FRS is a superconducting fragment separator, it will be able to separate all kinds of nuclear projectile fragments of primary heavy ion beams including Uranium with energies up to 1.5 GeV/u and intensities up to 3x1011 particles per second. During operation activation of several components, especially the production target and the beam catchers will take place. For handling of highly activated components it is foreseen to have a hot cell with connected storage place. All calculations for the optimisation of the shielding design of the SIS100, the Super-FRS and the hot cell were performed using the Monte Carlo code FLUKA, results are presented.

PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-10-22

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

Challenges dealing with depleted uranium in Germany - Reuse or disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moeller, Kai D.

2007-07-01

During enrichment large amounts of depleted Uranium are produced. In Germany every year 2.800 tons of depleted uranium are generated. In Germany depleted uranium is not classified as radioactive waste but a resource for further enrichment. Therefore since 1996 depleted Uranium is sent to ROSATOM in Russia. However it still has to be dealt with the second generation of depleted Uranium. To evaluate the alternative actions in case a solution has to be found in Germany, several studies have been initiated by the Federal Ministry of the Environment. The work that has been carried out evaluated various possibilities to dealmore » with depleted uranium. The international studies on this field and the situation in Germany have been analyzed. In case no further enrichment is planned the depleted uranium has to be stored. In the enrichment process UF{sub 6} is generated. It is an international consensus that for storage it should be converted to U{sub 3}O{sub 8}. The necessary technique is well established. If the depleted Uranium would have to be characterized as radioactive waste, a final disposal would become necessary. For the planned Konrad repository - a repository for non heat generating radioactive waste - the amount of Uranium is limited by the licensing authority. The existing license would not allow the final disposal of large amounts of depleted Uranium in the Konrad repository. The potential effect on the safety case has not been roughly analyzed. As a result it may be necessary to think about alternatives. Several possibilities for the use of depleted uranium in the industry have been identified. Studies indicate that the properties of Uranium would make it useful in some industrial fields. Nevertheless many practical and legal questions are open. One further option may be the use as shielding e.g. in casks for transport or disposal. Possible techniques for using depleted Uranium as shielding are the use of the metallic Uranium as well as the inclusion in concrete. Another possibility could be the use of depleted uranium for the blending of High enriched Uranium (HEU) or with Plutonium to MOX-elements. (authors)« less

Coating method for graphite

DOEpatents

Banker, John G.; Holcombe, Jr., Cressie E.

1977-01-01

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided comprising coating the graphite surface with a suspension of Y.sub.2 O.sub.3 particles in water containing about 1.5 to 4% by weight sodium carboxymethylcellulose.

Coating method for graphite

DOEpatents

Banker, J.G.; Holcombe, C.E. Jr.

1975-11-06

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided. The graphite surface is coated with a suspension of Y/sub 2/O/sub 3/ particles in water containing about 1.5 to 4 percent by weight sodium carboxymethylcellulose.

Aerosol Sampling System for Collection of Capstone Depleted Uranium Particles in a High-Energy Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holmes, Thomas D.; Guilmette, Raymond A.; Cheng, Yung-Sung

2009-03-01

The Capstone Depleted Uranium Aerosol Study was undertaken to obtain aerosol samples resulting from a kinetic-energy cartridge with a large-caliber depleted uranium (DU) penetrator striking an Abrams or Bradley test vehicle. The sampling strategy was designed to (1) optimize the performance of the samplers and maintain their integrity in the extreme environment created during perforation of an armored vehicle by a DU penetrator, (2) collect aerosols as a function of time post-impact, and (3) obtain size-classified samples for analysis of chemical composition, particle morphology, and solubility in lung fluid. This paper describes the experimental setup and sampling methodologies used tomore » achieve these objectives. Custom-designed arrays of sampling heads were secured to the inside of the target in locations approximating the breathing zones of the vehicle commander, loader, gunner, and driver. Each array was designed to support nine filter cassettes and nine cascade impactors mounted with quick-disconnect fittings. Shielding and sampler placement strategies were used to minimize sampler loss caused by the penetrator impact and the resulting fragments of eroded penetrator and perforated armor. A cyclone train was used to collect larger quantities of DU aerosol for chemical composition and solubility. A moving filter sample was used to obtain semicontinuous samples for depleted uranium concentration determination. Control for the air samplers was provided by five remotely located valve control and pressure monitoring units located inside and around the test vehicle. These units were connected to a computer interface chassis and controlled using a customized LabVIEW engineering computer control program. The aerosol sampling arrays and control systems for the Capstone study provided the needed aerosol samples for physicochemical analysis, and the resultant data were used for risk assessment of exposure to DU aerosol.« less

TREATMENT OF URANIUM SURFACES

DOEpatents

Slunder, C.J.

1959-02-01

An improved process is presented for prcparation of uranium surfaces prior to electroplating. The surfacc of the uranium to be electroplated is anodized in a bath comprising a solution of approximately 20 to 602 by weight of phosphoric acid which contains about 20 cc per liter of concentrated hydrochloric acid. Anodization is carried out for approximately 20 minutes at a current density of about 0.5 amperes per square inch at a temperature of about 35 to 45 C. The oxidic film produced by anodization is removed by dipping in strong nitric acid, followed by rinsing with water just prior to electroplating.

Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

NASA Astrophysics Data System (ADS)

Canion, Bonnie; McConchie, Seth; Landsberger, Sheldon

2017-07-01

This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

Routine inspection effort required for verification of a nuclear material production cutoff convention

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dougherty, D.; Fainberg, A.; Sanborn, J.

On 27 September 1993, President Clinton proposed {open_quotes}... a multilateral convention prohibiting the production of highly enriched uranium or plutonium for nuclear explosives purposes or outside of international safeguards.{close_quotes} The UN General Assembly subsequently adopted a resolution recommending negotiation of a non-discriminatory, multilateral, and internationally and effectively verifiable treaty (hereinafter referred to as {open_quotes}the Cutoff Convention{close_quotes}) banning the production of fissile material for nuclear weapons. The matter is now on the agenda of the Conference on Disarmament, although not yet under negotiation. This accord would, in effect, place all fissile material (defined as highly enriched uranium and plutonium) produced aftermore » entry into force (EIF) of the accord under international safeguards. {open_quotes}Production{close_quotes} would mean separation of the material in question from radioactive fission products, as in spent fuel reprocessing, or enrichment of uranium above the 20% level, which defines highly enriched uranium (HEU). Facilities where such production could occur would be safeguarded to verify that either such production is not occurring or that all material produced at these facilities is maintained under safeguards.« less

Size-dependent microstructures in rapidly solidified uranium-niobium powder particles

DOE PAGES

McKeown, Joseph T.; Hsiung, Luke L.; Park, Jong M.; ...

2016-06-14

The microstructures of rapidly solidified U-6wt%Nb powder particles synthesized by centrifugal atomization were characterized using scanning electron microscopy and transmission electron microscopy. Observed variations in microstructure are related to particle sizes. All of the powder particles exhibited a two-zone microstructure. The formation of this two-zone microstructure is described by a transition from solidification controlled by internal heat flow and high solidification rate during recalescence (micro-segregation-free or partitionless growth) to solidification controlled by external heat flow with slower solidification rates (dendritic growth with solute redistribution). The extent of partitionless solidification increased with decreasing particle size due to larger undercoolings in smallermore » particles prior to solidification. The metastable phases that formed are related to variations in Nb concentration across the particles. Lastly, the microstructures of the powders were heavily twinned.« less

PROCESS FOR PRODUCING URANIUM HALIDES

DOEpatents

Murphree, E.V.

1957-10-29

A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

Manufacturing Process Development to Produce Depleted Uranium Wire for EBAM Feedstock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexander, David John; Clarke, Kester Diederik; Coughlin, Daniel Robert

2015-06-30

Wire produced from depleted uranium (DU) is needed as feedstock for the Electron-Beam Additive Manufacturing (EBAM) process. The goal is to produce long lengths of DU wire with round or rectangular cross section, nominally 1.5 mm (0.060 inches). It was found that rolling methods, rather than swaging or drawing, are preferable for production of intermediate quantities of DU wire. Trials with grooveless rolling have shown that it is suitable for initial reductions of large stock. Initial trials with grooved rolling have been successful, for certain materials. Modified square grooves (square round-bottom vee grooves) with 12.5 % reduction of area permore » pass have been selected for the reduction process.« less

Uranium deposits of Brazil

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

Brazil is a country of vast natural resources, including numerous uranium deposits. In support of the country`s nuclear power program, Brazil has developed the most active uranium industry in South America. Brazil has one operating reactor (Angra 1, a 626-MWe PWR), and two under construction. The country`s economic challenges have slowed the progress of its nuclear program. At present, the Pocos de Caldas district is the only active uranium production. In 1990, the Cercado open-pit mine produced approximately 45 metric tons (MT) U{sub 3}O{sub 8} (100 thousand pounds). Brazil`s state-owned uranium production and processing company, Uranio do Brasil, announced itmore » has decided to begin shifting its production from the high-cost and nearly depleted deposits at Pocos de Caldas, to lower-cost reserves at Lagoa Real. Production at Lagoa Real is schedules to begin by 1993. In addition to these two districts, Brazil has many other known uranium deposits, and as a whole, it is estimated that Brazil has over 275,000 MT U{sub 3}O{sub 8} (600 million pounds U{sub 3}O{sub 8}) in reserves.« less

Solar wind alpha particle capture at Mars and Venus

NASA Astrophysics Data System (ADS)

Stenberg, Gabriella; Barabash, Stas; Nilsson, Hans; Fedorov, A.; Brain, David; André, Mats

Helium is detected in the atmospheres of both Mars and Venus. It is believed that radioactive decay of uranium and thorium in the interior of the planets' is not sufficient to account for the abundance of helium observed. Alpha particles in the solar wind are suggested to be an additional source of helium, especially at Mars. Recent hybrid simulations show that as much as 30We use ion data from the ASPERA-3 and ASPERA-4 instruments on Mars and Venus Express to estimate how efficient solar wind alpha particles are captured in the atmospheres of the two planets.

Monitoring genotoxic exposure in uranium mines.

PubMed Central

Srám, R J; Dobiás, L; Rössner, P; Veselá, D; Veselý, D; Rakusová, R; Rericha, V

1993-01-01

Recent data from deep uranium mines in Czechoslovakia indicated that mines are exposed to other mutagenic factors in addition to radon daughter products. Mycotoxins were identified as a possible source of mutagens in these mines. Mycotoxins were examined in 38 samples from mines and in throat swabs taken from 116 miners and 78 controls. The following mycotoxins were identified from mines samples: aflatoxins B1 and G1, citrinin, citreoviridin, mycophenolic acid, and sterigmatocystin. Some mold strains isolated from mines and throat swabs were investigated for mutagenic activity by the SOS chromotest and Salmonella assay with strains TA100 and TA98. Mutagenicity was observed, especially with metabolic activation in vitro. These data suggest that mycotoxins produced by molds in uranium mines are a new genotoxic factor for uranium miners. PMID:8143610

Leachability of uranium and other elements from freshly erupted volcanic ash

USGS Publications Warehouse

Smith, D.B.; Zielinski, R.A.; Rose, W.I.

1982-01-01

A study of leaching of freshly erupted basaltic and dacitic air-fall ash and bomb fragment samples, unaffected by rain, shows that glass dissolution is the dominant process by which uranium is initially mobilized from air-fall volcanic ash. Si, Li, and V are also preferentially mobilized by glass dissolution. Gaseous transfer followed by fixation of soluble uranium species on volcanic-ash particles is not an important process affecting uranium mobility. Gaseous transfer, however, may be important in forming water-soluble phases, adsorbed to ash surfaces, enriched in the economically and environmentally important elements Zn, Cu, Cd, Pb, B, F, and Ba. Quick removal of these adsorbed elements by the first exposure of freshly erupted ash to rain and surface water may pose short-term hazards to certain forms of aquatic and terrestrial life. Such rapid release of material may also represent the first step in transportation of economically important elements to environments favorable for precipitation into deposits of commercial interest. Ash samples collected from the active Guatemalan volcanoes Fuego and Pacaya (high-Al basalts) and Santiaguito (hornblende-hypersthene dacite); bomb fragments from Augustine volcano (andesite-dacite), Alaska, and Heimaey (basalt), Vestmann Islands, Iceland; and fragments of "rhyolitic" pumice from various historic eruptions were subjected to three successive leaches with a constant water-to-ash weight ratio of 4:1. The volcanic material was successively leached by: (1) distilled-deionized water (pH = 5.0-5.5) at room temperature for 24 h, which removes water-soluble gases and salts adsorbed on ash surfaces during eruption; (2) dilute HCl solution (pH = 3.5-4.0) at room temperature for 24 h, which continues the attack initiated by the water and also attacks acid-soluble sulfides and oxides; (3) a solution 0.05 M in both Na,CO, and NaHCO, (pH = 9.9) at 80°C for one week, which preferentially dissolves volcanic glass. The first two leaches mimic interaction of ash with rain produced in the vicinity of an active eruption. The third leach accelerates the effect of prolonged contact of volcanic ash with alkaline ground water present during ash diagenesis.

40 CFR 141.55 - Maximum contaminant level goals for radionuclides.

Code of Federal Regulations, 2013 CFR

2013-07-01

... radionuclides. 141.55 Section 141.55 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) WATER PROGRAMS (CONTINUED) NATIONAL PRIMARY DRINKING WATER REGULATIONS Maximum Contaminant Level Goals and... and radium-228 Zero. 2. Gross alpha particle activity (excluding radon and uranium) Zero. 3. Beta...

40 CFR 141.55 - Maximum contaminant level goals for radionuclides.

Code of Federal Regulations, 2012 CFR

2012-07-01

... radionuclides. 141.55 Section 141.55 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) WATER PROGRAMS (CONTINUED) NATIONAL PRIMARY DRINKING WATER REGULATIONS Maximum Contaminant Level Goals and... and radium-228 Zero. 2. Gross alpha particle activity (excluding radon and uranium) Zero. 3. Beta...

40 CFR 141.55 - Maximum contaminant level goals for radionuclides.

Code of Federal Regulations, 2014 CFR

2014-07-01

... radionuclides. 141.55 Section 141.55 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) WATER PROGRAMS (CONTINUED) NATIONAL PRIMARY DRINKING WATER REGULATIONS Maximum Contaminant Level Goals and... and radium-228 Zero. 2. Gross alpha particle activity (excluding radon and uranium) Zero. 3. Beta...

Cost Analysis of Remediation Systems for Depleted Uranium

DTIC Science & Technology

2014-04-01

situ stabilization and phytoremediation . In-situ stabilization includes the use of amend- ments, capping, and grouting to immobilize the contaminant in...its current location. Adding amendments to the soil solidifies the DU into insoluble particles. Phytoremediation is the use of plants to help reduce

Characterization of uranium carbide target materials to produce neutron-rich radioactive beams

NASA Astrophysics Data System (ADS)

Tusseau-Nenez, Sandrine; Roussière, Brigitte; Barré-Boscher, Nicole; Gottberg, Alexander; Corradetti, Stefano; Andrighetto, Alberto; Cheikh Mhamed, Maher; Essabaa, Saïd; Franberg-Delahaye, Hanna; Grinyer, Joanna; Joanny, Loïc; Lau, Christophe; Le Lannic, Joseph; Raynaud, Marc; Saïd, Abdelhakim; Stora, Thierry; Tougait, Olivier

2016-03-01

In the framework of a R&D program aiming to develop uranium carbide (UCx) targets for radioactive nuclear beams, the Institut de Physique Nucléaire d'Orsay (IPNO) has developed an experimental setup to characterize the release of various fission fragments from UCx samples at high temperature. The results obtained in a previous study have demonstrated the feasibility of the method and started to correlate the structural properties of the samples and their behavior in terms of nuclear reaction product release. In the present study, seven UCx samples have been systematically characterized in order to better understand the correlation between their physicochemical characteristics and release properties. Two very different samples, the first one composed of dense UC and the second one of highly porous UCx made of multi-wall carbon nanotubes, were provided by the ActILab (ENSAR) collaboration. The others were synthesized at IPNO. The systems for irradiation and heating necessary for the release studies have been improved with respect to those used in previous studies. The results show that the open porosity is hardly the limiting factor for the fission product release. The homogeneity of the microstructure and the pore size distribution contributes significantly to the increase of the release. The use of carbon nanotubes in place of traditional micrometric graphite particles appears to be promising, even if the homogeneity of the microstructure can still be enhanced.

Analysis of fission and activation radionuclides produced by a uranium-fueled nuclear detonation and identification of the top dose-producing radionuclides.

PubMed

Kraus, Terry; Foster, Kevin

2014-08-01

The radiological assessment of the nuclear fallout (i.e., fission and neutron-activation radionuclides) from a nuclear detonation is complicated by the large number of fallout radionuclides. This paper provides the initial isotopic source term inventory of the fallout from a uranium-fueled nuclear detonation and identifies the significant and insignificant radiological dose producing radionuclides over 11 dose integration time periods (time phases) of interest. A primary goal of this work is to produce a set of consistent, time phase-dependent lists of the top dose-producing radionuclides that can be used to prepare radiological assessment calculations and data products (e.g., maps of areas that exceed protective action guidelines) in support of public and worker protection decisions. The ranked lists of top dose-producing radionuclides enable assessors to perform atmospheric dispersion modeling and radiological dose assessment modeling more quickly by using relatively short lists of radionuclides without significantly compromising the accuracy of the modeling and the dose projections. This paper also provides a superset-list of the top dose-producing fallout radionuclides from a uranium-fueled nuclear detonation that can be used to perform radiological assessments over any desired time phase. Furthermore, this paper provides information that may be useful to monitoring and sampling and laboratory analysis personnel to help understand which radionuclides are of primary concern. Finally, this paper may be useful to public protection decision makers because it shows the importance of quickly initiating public protection actions to minimize the radiological dose from fallout.

Comparison of actinide production in traveling wave and pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

Uranium redox transition pathways in acetate-amended sediments

USGS Publications Warehouse

Bargar, John R.; Williams, Kenneth H.; Campbell, Kate M.; Long, Philip E.; Stubbs, Joanne E.; Suvorova, Elenal I.; Lezama-Pacheco, Juan S.; Alessi, Daniel S.; Stylo, Malgorzata; Webb, Samuel M.; Davis, James A.; Giammar, Daniel E.; Blue, Lisa Y.; Bernier-Latmani, Rizlan

2013-01-01

Redox transitions of uranium [from U(VI) to U(IV)] in low-temperature sediments govern the mobility of uranium in the environment and the accumulation of uranium in ore bodies, and inform our understanding of Earth’s geochemical history. The molecular-scale mechanistic pathways of these transitions determine the U(IV) products formed, thus influencing uranium isotope fractionation, reoxidation, and transport in sediments. Studies that improve our understanding of these pathways have the potential to substantially advance process understanding across a number of earth sciences disciplines. Detailed mechanistic information regarding uranium redox transitions in field sediments is largely nonexistent, owing to the difficulty of directly observing molecular-scale processes in the subsurface and the compositional/physical complexity of subsurface systems. Here, we present results from an in situ study of uranium redox transitions occurring in aquifer sediments under sulfate-reducing conditions. Based on molecular-scale spectroscopic, pore-scale geochemical, and macroscale aqueous evidence, we propose a biotic–abiotic transition pathway in which biomass-hosted mackinawite (FeS) is an electron source to reduce U(VI) to U(IV), which subsequently reacts with biomass to produce monomeric U(IV) species. A species resembling nanoscale uraninite is also present, implying the operation of at least two redox transition pathways. The presence of multiple pathways in low-temperature sediments unifies apparently contrasting prior observations and helps to explain sustained uranium reduction under disparate biogeochemical conditions. These findings have direct implications for our understanding of uranium bioremediation, ore formation, and global geochemical processes.

Hydrogeochemical and stream sediment detailed geochemical survey for Edgemont, South Dakota; Wyoming

DOE Office of Scientific and Technical Information (OSTI.GOV)

Butz, T.R.; Dean, N.E.; Bard, C.S.

1980-05-31

Results of the Edgemont detailed geochemical survey are reported. Field and laboratory data are presented for 109 groundwater and 419 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are given. A generalized geologic map of the survey area is provided, and pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. Groundwaters containing greater than or equal to 7.35 ppB uranium are present in scattered clusters throughout the area sampled. Most of these groundwaters are from wells drilled where the Inyan Kara Group is exposed at themore » surface. The exceptions are a group of samples in the northwestern part of the area sampled and south of the Dewey Terrace. These groundwaters are also produced from the Inyan Kara Group where it is overlain by the Graneros Group and alluvium. The high uranium groundwaters along and to the south of the terrace are characterized by high molybdenum, uranium/specific conductance, and uranium/sulfate values. Many of the groundwaters sampled along the outcrop of the Inyan Kara Group are near uranium mines. Groundwaters have high amounts of uranium and molybdenum. Samples taken downdip are sulfide waters with low values of uranium and high values of arsenic, molybdenum, selenium, and vanadium. Stream sediments containing greater than or equal to 5.50 ppM soluble uranium are concentrated in basins draining the Graneros and Inyan Kara Groups. These values are associated with high values for arsenic, selenium, and vanadium in samples from both groups. Anomalous values for these elements in the Graneros Group may be caused by bentonite beds contained in the rock units. As shown on the geochemical distribution plot, high uranium values that are located in the Inyan Kara Group are almost exclusively draining open-pit uranium mines.« less

The roles of organic matter in the formation of uranium deposits in sedimentary rocks

USGS Publications Warehouse

Spirakis, C.S.

1996-01-01

Because reduced uranium species have a much smaller solubility than oxidized uranium species and because of the strong association of organic matter (a powerful reductant) with many uranium ores, reduction has long been considered to be the precipitation mechanism for many types of uranium deposits. Organic matter may also be involved in the alterations in and around tabular uranium deposits, including dolomite precipitation, formation of silicified layers, iron-titanium oxide destruction, dissolution of quartz grains, and precipitation of clay minerals. The diagenetic processes that produced these alterations also consumed organic matter. Consequently, those tabular deposits that underwent the more advanced stages of diagenesis, including methanogenesis and organic acid generation, display the greatest range of alterations and contain the smallest amount of organic matter. Because of certain similarities between tabular uranium deposits and Precambrian unconformity-related deposits, some of the same processes might have been involved in the genesis of Precambrian unconformity-related deposits. Hydrologic studies place important constraints on genetic models of various types of uranium deposits. In roll-front deposits, oxidized waters carried uranium to reductants (organic matter and pyrite derived from sulfate reduction by organic matter). After these reductants were oxidized at any point in the host sandstone, uranium minerals were reoxidized and transported further down the flow path to react with additional reductants. In this manner, the uranium ore migrated through the sandstone at a rate slower than the mineralizing ground water. In the case of tabular uranium deposits, the recharge of surface water into the ground water during flooding of lakes carried soluble humic material to the water table or to an interface where humate precipitated in tabular layers. These humate layers then established the chemical conditions for mineralization and related alterations. In the case of Precambrian unconformity-related deposits, free thermal convection in the thick sandstones overlying the basement rocks carried uranium to concentrations of organic matter in the basement rocks.

Different pattern of brain pro-/anti-oxidant activity between depleted and enriched uranium in chronically exposed rats.

PubMed

Lestaevel, P; Romero, E; Dhieux, B; Ben Soussan, H; Berradi, H; Dublineau, I; Voisin, P; Gourmelon, P

2009-04-05

Uranium is not only a heavy metal but also an alpha particle emitter. The main toxicity of uranium is expected to be due to chemiotoxicity rather than to radiotoxicity. Some studies have demonstrated that uranium induced some neurological disturbances, but without clear explanations. A possible mechanism of this neurotoxicity could be the oxidative stress induced by reactive oxygen species imbalance. The aim of the present study was to determine whether a chronic ingestion of uranium induced anti-oxidative defence mechanisms in the brain of rats. Rats received depleted (DU) or 4% enriched (EU) uranyl nitrate in the drinking water at 2mg(-1)kg(-1)day(-1) for 9 months. Cerebral cortex analyses were made by measuring mRNA and protein levels and enzymatic activities. Lipid peroxidation, an oxidative stress marker, was significantly enhanced after EU exposure, but not after DU. The gene expression or activity of the main antioxidant enzymes, i.e. superoxide dismutase (SOD), catalase (CAT) and glutathione peroxidase (GPx), increased significantly after chronic exposure to DU. On the contrary, oral EU administration induced a decrease of these antioxidant enzymes. The NO-ergic pathway was almost not perturbed by DU or EU exposure. Finally, DU exposure increased significantly the transporters (Divalent-Metal-Transporter1; DMT1), the storage molecule (ferritin) and the ferroxidase enzyme (ceruloplasmin), but not EU. These results illustrate that oxidative stress plays a key role in the mechanism of uranium neurotoxicity. They showed that chronic exposure to DU, but not EU, seems to induce an increase of several antioxidant agents in order to counteract the oxidative stress. Finally, these results demonstrate the importance of the double toxicity, chemical and radiological, of uranium.

Study on the Classification of the Uranium Mineral of Venta de Cardena; ESTUDIO SOBRE LA CLASIFICACION DEL MINERAL DE VENTA DE CARDENA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Otero, A.R.

1959-01-01

The behavior of uranium mineral from Venta de Cardena in a spiral classifier which operates in a closed system with a ball mill was studied to obtain data for the design of a milling-classification system with a production capacity of 200 tons per day with a particle size less than 0.417 mm. The characteristics of such a system, the problems in normal operation, the inconveriences which these cause, and their solution were investigated. Correlations between these tests and the results obtained with long glass tubes are presented. (J.S.R.)

Ballistic piston fissioning plasma experiment.

NASA Technical Reports Server (NTRS)

Miller, B. E.; Schneider, R. T.; Thom, K.; Lalos, G. T.

1971-01-01

The production of fissioning uranium plasma samples such that the fission fragment stopping distance is less than the dimensions of the plasma is approached by using a ballistic piston device for the compression of uranium hexafluoride. The experimental apparatus is described. At room temperature the gun can be loaded up to 100 torr UF6 partial pressure, but at compression a thousand fold increase of pressure can be obtained at a particle density on the order of 10 to the 19th power per cu cm. Limited spectral studies of UF6 were performed while obtaining the pressure-volume data. The results obtained and their implications are discussed.

METAL EXTRACTION PROCESS

DOEpatents

Lewis, G.W. Jr.; Rhodes, D.E.

1957-11-01

An improved method for extracting uranium from aqueous solutions by solvent extraction is presented. A difficulty encountered in solvent extraction operations using an organic extractant (e.g., tributyl phosphate dissolved in kerosene or carbon tetrachloride) is that emulsions sometimes form, and phase separation is difficult or impossible. This difficulty is overcome by dissolving the organic extractant in a molten wax which is a solid at operating temperatures. After cooling, the wax which now contains the extractant, is broken into small particles (preferably flakes) and this wax complex'' is used to contact the uranium bearing solutions and extract the metal therefrom. Microcrystalline petroleum wax and certain ethylene polymers have been found suitable for this purpose.

The Permo-Triassic uranium deposits of Gondwanaland

NASA Astrophysics Data System (ADS)

le Roux, J. P.; Toens, P. D.

The world's uranium provinces are time bound and occur in five distinct periods ranging from the Proterozoic to the Recent. One of these periods embraces the time of Gondwana sedimentation and probably is related to the proliferation of land plants from the Devonian on-ward. Decaying vegetal matter produced reducing conditions that enhanced uranium precipitation. The association of uranium with molassic basins adjacent to uplifted granitic and volcanic arcs suggests that lithospheric plate subduction, leading to anatexis of basement rocks and andesitic volcanism, created favorable conditions for uranium mineralization. Uranium occurrences of Gondwana age are of four main types: sandstone-hosted, coal-hosted, pelite-hosted, and vein-type deposits. Sandstone-hosted deposits commonly occur in fluviodeltaic sediments and are related to the presence of organic matter. These deposits commonly are enriched in molybdenum and other base metal sulfides and have been found in South Africa, Zimbabwe, Zambia, Angola, Niger, Madagascar, India, Australia, Argentina, and Brazil. Coalhosted deposits contain large reserves of uranium but are of low grade. In Africa they are mostly within the Permian Ecca Group and its lateral equivalents, as in the Springbok Flats, Limpopo, Botswana, and Tanzania basins. Uraniferous black shales are present in the Gabon and Amazon basins but grades are low. Vein-type uranium is found in Argentina, where it occurs in clustered veins crosscutting sedimentary rocks and quartz porphyries.

Simulations of Multi-Gamma Coincidences From Neutron-Induced Fission in Special Nuclear Materials

NASA Astrophysics Data System (ADS)

Kane, Steven; Gozani, Tsahi; King, Michael J.; Kwong, John; Brown, Craig; Gary, Charles; Firestone, Murray I.; Nikkel, James A.; McKinsey, Daniel N.

2013-04-01

A study is presented on the detection of illicit special nuclear materials (SNM) in cargo containers using a conceptual neutron-based inspection system with xenon-doped liquefied argon (LAr(Xe)) scintillation detectors for coincidence gamma-ray detection. For robustness, the system is envisioned to exploit all fission signatures, namely both prompt and delayed neutron and gamma emissions from fission reactions induced in SNM. However, this paper focuses exclusively on the analysis of the prompt gamma ray emissions. The inspection system probes a container using neutrons produced either by (d, D) or (d, T) in pulsed form or from an associated particle neutron generator to exploit the associated particle imaging (API) technique, thereby achieving background reduction and imaging. Simulated signal and background estimates were obtained in MCNPX (2.7) for a 2 kg sphere of enriched uranium positioned at the center of a 1m × 1m × 1m container, which is filled uniformly with wood or iron cargos at 0.1 g/cc or 0.4 g/cc. Detection time estimates are reported assuming probabilities of detection of 95% and false alarm of 0.5%.

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2010-08-16

nuclear weapons facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near...brief interruptions.”85 84 Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds...confiscated a rare metal used to produce alloy steel (called vanadium) being smuggled to North Korea. In the same month, China’s NHI Shenyang Mining

The Ames Project (1942-1946)

ScienceCinema

None

2018-04-26

The Ames Laboratory was officially founded on May 17, 1947, following development of a process to purify uranium metal for the historic Manhattan Project. From 1942 to 1946, Ames Lab scientists produced over two-million pounds of uranium metal. A U.S. Department of Energy national research laboratory, the Ames Laboratory creates materials and energy solutions. Iowa State University operates Ames Laboratory under contract with the DOE.

Absolute intensity of radiation emitted by uranium plasmas

NASA Technical Reports Server (NTRS)

Jalufka, N. W.; Lee, J. H.; Mcfarland, D. R.

1975-01-01

The absolute intensity of radiation emitted by fissioning and nonfissioning uranium plasmas in the spectral range from 350 nm to 1000 nm was measured. The plasma was produced in a plasma-focus apparatus and the plasma properties are simular to those anticipated for plasma-core nuclear reactors. The results are expected to contribute to the establishment of design criteria for the development of plasma-core reactors.

Production and Evaluation of 236gNp and Reference Materials for Naturally Occurring Radioactive Materials

NASA Astrophysics Data System (ADS)

Larijani, Cyrus Kouroush

This thesis is based on the development of a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. The isobaric distribution of fission residues produced following the bombardment of a natural uranium target with a beam of 25 MeV protons has been evaluated. Decay analysis of thirteen isobarically distinct fission residues were carried out using high-resolution gamma-ray spectrometry at the UK National Physical Laboratory. Stoichiometric abundances were calculated via the determination of absolute activity concentrations associated with the longest-lived members of each isobaric chain. This technique was validated by computational modelling of likely sequential decay processes through an isobaric decay chain. The results were largely in agreement with previously published values for neutron bombardments on natural uranium at energies of 14 MeV. Higher relative yields of products with mass numbers A 110-130 were found, consistent with the increasing yield of these radionuclides as the bombarding energy is increased. Using literature values for the production cross-section for fusion of protons with uranium targets, it is estimated that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution alpha and gamma-ray spectrometry. In a separate research theme, reliable measurement of Naturally Occurring Radioactive Materials is of significance in order to comply with environmental regulations and for radiological protection purposes. The thesis describes the standardisation of three reference materials, namely Sand, Tuff and TiO2 which can serve as quality control materials to achieve traceability, method validation and instrument calibration. The sample preparation, material characterization via gamma, alpha and Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and the assignment of values for both the 4n Thorium and 4n + 2 Uranium decay series are presented.

Study on kinetics of adsorption of humic acid modified by ferric chloride on U(VI)

NASA Astrophysics Data System (ADS)

Zhang, Y. Y.; Lv, J. W.; Song, Y.; Dong, X. J.; Fang, Q.

2017-11-01

In order to reveal the adsorption mechanism of the ferric chloride modified humic acid on uranium, the influence of pH value and contact time of adsorption on uranium was studied through a series of batch experiments. Meanwhile, the adsorption kinetics was analyzed with pseudo-first order kinetic model and pseudo-second order kinetic model. The results show that adsorption is affected by the pH value of the solution and by contract time, and the best condition for adsorption on uranium is at pH=5 and the adsorption equilibrium time is about 80 min. Kinetics of HA-Fe adsorption on uranium accords with pseudo-second order kinetic model. The adsorption is mainly chemical adsorption, and complexes were produced by the reaction between uranium ions and the functional groups on the surface of HA-Fe, which can provide reference for further study of humic acid effecting on the migration of U(VI) in soil.

In situ remediation of uranium contaminated groundwater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dwyer, B.P.; Marozas, D.C.

1997-02-01

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptablemore » regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field results are discussed with regard to other potential contaminated groundwater treatment applications.« less

In situ remediation of uranium contaminated groundwater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dwyer, B.P.; Marozas, D.C.

1997-12-31

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment - various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ tomore » acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field preliminary results are discussed with regard to other potential contaminated groundwater treatment applications.« less

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rao, Linfeng

A literature survey has been conducted to collect information on the International R&D activities in the extraction of uranium from seawater for the period from the 1960s till the year of 2010. The reported activities, on both the laboratory scale bench experiments and the large scale marine experiments, were summarized by country/region in this report. Among all countries where such activities have been reported, Japan has carried out the most advanced large scale marine experiments with the amidoxime-based system, and achieved the collection efficiency (1.5 g-U/kg-adsorbent for 30 days soaking in the ocean) that could justify the development of industrialmore » scale marine systems to produce uranium from seawater at the price competitive with those from conventional uranium resources. R&D opportunities are discussed for improving the system performance (selectivity for uranium, loading capacity, chemical stability and mechanical durability in the sorption-elution cycle, and sorption kinetics) and making the collection of uranium from seawater more economically competitive.« less

Characterization and Testing of Amidoxime-Based Adsorbent Materials to Extract Uranium from Natural Seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuo, Li-Jung; Janke, Christopher J.; Wood, Jordana R.

Extraction of uranium (U) from seawater for use as a nuclear fuel is a significant challenge due to the low concentration of U in seawater (~3.3 ppb) and difficulties to selectively extract U from the background of major and trace elements in seawater. The Pacific Northwest National Laboratory (PNNL)’s Marine Sciences Laboratory (MSL) has been serving as a marine test site for determining performance characteristics (adsorption capacity, adsorption kinetics, and selectivity) of novel amidoxime-based polymeric adsorbents developed at Oak Ridge National Laboratory (ORNL) under natural seawater exposure conditions. This manuscript describes the performance of three formulations (38H, AF1, AI8) ofmore » amidoxime-based polymeric adsorbents produced at ORNL in MSL’s ambient seawater testing facility. The adsorbents were produced in two forms, fibrous material (40-100 mg samples) and braided material (5-10 g samples), and exposed to natural seawater using flow-through columns and recirculating flumes. All three formulations demonstrated high 56 day uranium adsorption capacity (>3 g U/kg adsorbent). The AF1 formulation had the best uranium adsorption performance, with a 56 day capacity of 3.9 g U/ kg adsorbent, a saturation capacity of 5.4 g U/kg adsorbent, and ~25 days half-saturation time. The two exposure methods, flowthrough columns and flumes, were demonstrated to produce similar performance results, providing confidence that the test methods were reliable, that scaling up from 10’s of mg quantities of exposure in flow-through columns to gram quantities in flumes produced similar results, and confirm that the manufacturing process produces a homogeneous adsorbent. Adsorption kinetics appear to be element specific, with half-saturation times ranging from minutes for the major cations in seawater, to 8-10 weeks for V and Fe. Reducing the exposure time provides a potential pathway to improve the adsorption capacity of U by reducing the V/U ratio on the adsorbent.« less

Characterization and testing of amidoxime-based adsorbent materials to extract uranium from natural seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuo, Li-Jung; Janke, Christopher James; Wood, Jordana

Extraction of uranium (U) from seawater for use as a nuclear fuel is a significant challenge due to the low concentration of U in seawater (~3.3 ppb) and difficulties to selectively extract U from the background of major and trace elements in seawater. The Pacific Northwest National Laboratory (PNNL) s Marine Sciences Laboratory (MSL) has been serving as a marine test site for determining performance characteristics (adsorption capacity, adsorption kinetics, and selectivity) of novel amidoxime-based polymeric adsorbents developed at Oak Ridge National Laboratory (ORNL) under natural seawater exposure conditions. This report describes the performance of three formulations (38H, AF1, AI8)more » of amidoxime-based polymeric adsorbent produced at ORNL in MSL s ambient seawater testing facility. The adsorbents were produced in two forms, fibrous material (40-100 mg samples) and braided material (5-10 g samples), exposed to natural seawater using flow-through columns and recirculating flumes. All three formulations demonstrated high 56 day uranium adsorption capacity (>3 gU/kg adsorbent). The AF1 formulation had the best uranium adsorption performance, with 56-day capacity of 3.9 g U/kg adsorbent, saturation capacity of 5.4 g U/kg adsorbent, and ~25 days half-saturation time. The two exposure methods, flow-through columns and flumes were demonstrated to produce similar performance results, providing confidence that the test methods were reliable, that scaling up from 10 s of mg quantities of exposure in flow-through columns to gram quantities in flumes produced similar results, and that the manufacturing process produces a homogenous adsorbent. Adsorption kinetics appear to be element specific, with half-saturation times ranging from minutes for the major cations in seawater to 8-10weeks for V and Fe. Reducing the exposure time provides a potential pathway to improve the adsorption capacity of U by reducing the V/U ratio on the adsorbent.« less

Characterization and testing of amidoxime-based adsorbent materials to extract uranium from natural seawater

DOE PAGES

Kuo, Li-Jung; Janke, Christopher James; Wood, Jordana; ...

2015-11-19

Extraction of uranium (U) from seawater for use as a nuclear fuel is a significant challenge due to the low concentration of U in seawater (~3.3 ppb) and difficulties to selectively extract U from the background of major and trace elements in seawater. The Pacific Northwest National Laboratory (PNNL) s Marine Sciences Laboratory (MSL) has been serving as a marine test site for determining performance characteristics (adsorption capacity, adsorption kinetics, and selectivity) of novel amidoxime-based polymeric adsorbents developed at Oak Ridge National Laboratory (ORNL) under natural seawater exposure conditions. This report describes the performance of three formulations (38H, AF1, AI8)more » of amidoxime-based polymeric adsorbent produced at ORNL in MSL s ambient seawater testing facility. The adsorbents were produced in two forms, fibrous material (40-100 mg samples) and braided material (5-10 g samples), exposed to natural seawater using flow-through columns and recirculating flumes. All three formulations demonstrated high 56 day uranium adsorption capacity (>3 gU/kg adsorbent). The AF1 formulation had the best uranium adsorption performance, with 56-day capacity of 3.9 g U/kg adsorbent, saturation capacity of 5.4 g U/kg adsorbent, and ~25 days half-saturation time. The two exposure methods, flow-through columns and flumes were demonstrated to produce similar performance results, providing confidence that the test methods were reliable, that scaling up from 10 s of mg quantities of exposure in flow-through columns to gram quantities in flumes produced similar results, and that the manufacturing process produces a homogenous adsorbent. Adsorption kinetics appear to be element specific, with half-saturation times ranging from minutes for the major cations in seawater to 8-10weeks for V and Fe. Reducing the exposure time provides a potential pathway to improve the adsorption capacity of U by reducing the V/U ratio on the adsorbent.« less

Characterization of extreme ultraviolet laser ablation mass spectrometry for actinide trace analysis and nanoscale isotopic imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Green, Tyler; Kuznetsov, Ilya; Willingham, David

The purpose of this research was to characterize Extreme Ultraviolet Time-of-Flight (EUV TOF) Laser Ablation Mass Spectrometry for high spatial resolution elemental and isotopic analysis. We compare EUV TOF results with Secondary Ionization Mass Spectrometry (SIMS) to orient the EUV TOF method within the overall field of analytical mass spectrometry. Using the well-characterized NIST 61x glasses, we show that the EUV ionization approach produces relatively few molecular ion interferences in comparison to TOF SIMS. We demonstrate that the ratio of element ion to element oxide ion is adjustable with EUV laser pulse energy and that the EUV TOF instrument hasmore » a sample utilization efficiency of 0.014%. The EUV TOF system also achieves a lateral resolution of 80 nm and we demonstrate this lateral resolution with isotopic imaging of closely spaced particles or uranium isotopic standard materials.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scott, B.R.; Muggenburg, B.A.; Welsh, C.A.

The alpha emitter plutonium-238 ({sup 238}Pu), which is produced in uranium-fueled, light-water reactors, is used as a thermoelectric power source for space applications. Inhalation of a mixed oxide form of Pu is the most likely mode of exposure of workers and the general public. Occupational exposures to {sup 238}PuO{sub 2} have occurred in association with the fabrication of radioisotope thermoelectric generators. Organs and tissue at risk for deterministic and stochastic effects of {sup 238}Pu-alpha irradiation include the lung, liver, skeleton, and lymphatic tissue. Little has been reported about the effects of inhaled {sup 238}PuO{sub 2} on peripheral blood cell countsmore » in humans. The purpose of this study was to investigate hematological responses after a single inhalation exposure of Beagle dogs to alpha-emitting {sup 238}PuO{sub 2} particles and to extrapolate results to humans.« less

New insight into UO 2F 2 particulate structure by micro-Raman spectroscopy

DOE PAGES

Stefaniak, Elzbieta A.; Darchuk, Larysa; Sapundjiev, Danislav; ...

2013-02-19

Uranyl fluoride particles produced via hydrolysis of uranium hexafluoride have been deposited on different substrates: polished graphite disks, silver foil, stainless steel and gold-coated silicon wafer, and measured with micro-Raman spectroscopy (MRS). All three metallic substrates enhanced the Raman signal delivered by UO 2F 2 in comparison to graphite. The fundamental stretching of the U–O band appeared at 867 cm –1 in case of the graphite substrate, while in case of the others it was shifted to lower frequencies (down to 839 cm –1). All applied metallic substrates showed the expected effect of Raman signal enhancement; however the gold layermore » appeared to be most effective. Lastly, application of new substrates provides more information on the molecular structure of uranyl fluoride precipitation, which is interesting for nuclear safeguards and nuclear environmental analysis.« less

Failla Memorial Lecture: the future of heavy-ion science in biology and medicine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobias, C.A.

1985-07-01

An extensive review, with over 100 references, of the use of accelerator techniques in radiobiology is presented. Currently, beams of any stable isotope species up to uranium are available at kinetic energies of several hundred MeV/nucleon at the Berkeley Bevalac. The heavy ions hold interest for a broad spectrum of research because of their effectiveness in producing a series of major lesions in DNA along single particle tracks and because of the Bragg depth ionization properties that allow the precise deposition of highly localized doses deep in the human body. Heavy ions, when compared to low-LET radiation, have increased effectivenessmore » for mammalian cell lethality, chromosome mutations, and cell transformation. The molecular mechanisms are not completely understood but appear to involve fragmentation and reintegration of DNA. Heavy ions do not require the presence of oxygen for producing their effects. Heavy ions are effective in delaying or blocking the cell division process. These radiobiological properties, combined with the ability to deliver highly localized internal doses, make accelerated heavy ions potentially important radiotherapeutic tools. Other novel approaches include the utilization of radioactive heavy beams as instant tracers. Heavy-ion radiography and microscopy respond to delicate changes in tissue electron density. The authors laboratory is in the process of proposing a research biomedical heavy-ion accelerator; the availability of such machines would greatly accelerate cancer and brain research with particle beams.« less

Research and development on materials for the SPES target

NASA Astrophysics Data System (ADS)

Corradetti, Stefano; Andrighetto, Alberto; Manzolaro, Mattia; Scarpa, Daniele; Vasquez, Jesus; Rossignoli, Massimo; Monetti, Alberto; Calderolla, Michele; Prete, Gianfranco

2014-03-01

The SPES project at INFN-LNL (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro) is focused on the production of radioactive ion beams. The core of the SPES facility is constituted by the target, which will be irradiated with a 40 MeV, 200 µA proton beam in order to produce radioactive species. In order to efficiently produce and release isotopes, the material constituting the target should be able to work under extreme conditions (high vacuum and temperatures up to 2000 °C). Both neutron-rich and proton-rich isotopes will be produced; in the first case, carbon dispersed uranium carbide (UCx) will be used as a target, whereas to produce p-rich isotopes, several types of targets will have to be irradiated. The synthesis and characterization of different types of material will be reported. Moreover, the results of irradiation and isotopes release tests on different uranium carbide target prototypes will be discussed.

Monitoring genotoxic exposure in uranium mines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sram, R.J.; Vesela, D.; Vesely, D.

1993-10-01

Recent data from deep uranium mines in Czechoslovakia indicated that miners are exposed to other mutagenic factors in addition to radon daughter products. Mycotoxins were identified as a possible source of mutagens in these mines. Mycotoxins were examined in 38 samples from mines and in throat swabs taken from 116 miners and 78 controls. The following mycotoxins were identified from mines samples: aflatoxins B{sub 1} and G1, citrinin, citreoviridin, mycophenolic acid, and sterigmatocystin. Some mold strains isolated from mines and throat swabs were investigated for mutagenic activity by the SOS chromotest and Salmonella assay with strains TA100 and TA98. Mutagenicitymore » was observed, especially with metabolic activation in citro. These data suggest that mycotoxins produced by molds in uranium mines are a new genotoxic factor im uranium miners. 17 refs., 4 tabs.« less

Evaluating the risk from depleted uranium after the Boeing 747-258F crash in Amsterdam, 1992.

PubMed

Uijt de Haag, P A; Smetsers, R C; Witlox, H W; Krüs, H W; Eisenga, A H

2000-08-28

On 4 October 1992, a large cargo plane crashed into an apartment building in the Bijlmermeer quarter of Amsterdam. In the years following the accident, an increasing number of people started reporting health complaints, which they attributed to exposure to dangerous substances after the crash. Since the aircraft had been carrying depleted uranium as counterbalance weights and about 150 kg uranium had been found missing after clearance of the crash site, exposure to uranium oxide particles was pointed out as the possible cause of their health complaints. Six years after the accident, a risk analysis was therefore carried out to investigate whether the health complaints could be attributed to exposure to uranium oxide set free during the accident. The scientific challenge was to come up with reliable results, knowing that - considering the late date - virtually no data were available to validate any calculated result. The source term of uranium was estimated using both generic and specific data. Various dispersion models were applied in combination with the local setting and the meteorological conditions at the time of the accident to estimate the exposure of bystanders during the fire caused by the crash. Emphasis was given to analysing the input parameters, inter-comparing the various models and comparing model results with the scarce information available. Uranium oxide formed in the fire has a low solubility, making the chemical toxicity to humans less important than the radiotoxicity. Best-estimate results indicated that bystanders may have been exposed to a radiation dose of less than 1 microSv, whereas a worst-case approach indicated an upper limit of less than 1 mSv. This value is considerably less than the radiation dose for which acute effects are to be expected. It is therefore considered to be improbable that the missing uranium had indeed led to the health complaints reported.

Evolution of the radiation processing industry

NASA Astrophysics Data System (ADS)

Cleland, Marshall R.

2013-04-01

Early investigations of the effects of treating materials with ionizing radiations began in 1894 with the irradiation of gases at atmospheric pressure using cathode rays from a Crookes gas-discharge tube, in 1895 with the discovery of X-rays emitted from a Crookes tube, and in 1896 with the discovery of radioactivity in uranium. In 1897, small electrically charged particles were detected and identified in the gas discharges inside Crookes tubes. These particles were then named electrons. During the next three decades, it was found that these novel forms of energy could produce ions to initiate chemical reactions in some gases and liquids. By 1921, it had also been shown that insects, parasites and bacteria could be killed by treatment with ionizing radiation. In 1925, a high-vacuum tube with a thermionic cathode and a thin metallic anode was developed to produce electron beams in air by using accelerating potentials up to 250 kilovolts. That unique apparatus was the precursor of the many types of electron accelerators that have been developed since then for a variety of industrial applications. In 1929, the vulcanization of natural rubber without using any chemical additives was achieved by irradiation with electrons from a 250 kilovolt accelerator. In 1939, several liquid monomers were polymerized by treatment with gamma rays from radioactive nuclides. These early results were not exploited before the end of World War II because intense sources of ionizing radiation were not available then. Shortly after that war, there was increased interest in developing the peaceful uses of atomic energy, which included the chemical and biological effects of radiation exposures. Many uses that have been developed since then are described briefly in this paper. These industrial applications are now producing billions of US dollars in revenue every year.

Evolution of the radiation processing industry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cleland, Marshall R.

2013-04-19

Early investigations of the effects of treating materials with ionizing radiations began in 1894 with the irradiation of gases at atmospheric pressure using cathode rays from a Crookes gas-discharge tube, in 1895 with the discovery of X-rays emitted from a Crookes tube, and in 1896 with the discovery of radioactivity in uranium. In 1897, small electrically charged particles were detected and identified in the gas discharges inside Crookes tubes. These particles were then named electrons. During the next three decades, it was found that these novel forms of energy could produce ions to initiate chemical reactions in some gases andmore » liquids. By 1921, it had also been shown that insects, parasites and bacteria could be killed by treatment with ionizing radiation. In 1925, a high-vacuum tube with a thermionic cathode and a thin metallic anode was developed to produce electron beams in air by using accelerating potentials up to 250 kilovolts. That unique apparatus was the precursor of the many types of electron accelerators that have been developed since then for a variety of industrial applications. In 1929, the vulcanization of natural rubber without using any chemical additives was achieved by irradiation with electrons from a 250 kilovolt accelerator. In 1939, several liquid monomers were polymerized by treatment with gamma rays from radioactive nuclides. These early results were not exploited before the end of World War II because intense sources of ionizing radiation were not available then. Shortly after that war, there was increased interest in developing the peaceful uses of atomic energy, which included the chemical and biological effects of radiation exposures. Many uses that have been developed since then are described briefly in this paper. These industrial applications are now producing billions of US dollars in revenue every year.« less

Recovery of fissile materials from nuclear wastes

DOEpatents

Forsberg, Charles W.

1999-01-01

A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

Fertile-to-fissile and fission measurements for depleted uranium and thorium bombarded by 800-MeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilmore, J.S.; Russell, G.J.; Robinson, H.

Axial distributions of fissions and of fertile-to-fissile conversions in thick depleted uranium and thorium targets bombarded by 800-MeV protons have been measured. The amounts of /sup 239/Pu and /sup 233/U produced were determined by measuring the yields of /sup 239/Np and /sup 233/Pa, respectively. The number of fissions was deduced from fission product mass-yield curves. Integration of the axial distributions gave the total number of conversions and fissions occurring in the targets. For the uranium target, experimental results were 5.90 +- 0.25 fissions and 3.81 +- 0.01 atoms of /sup 239/Pu produced per incident portion. Corresponding calculated results were 6.14more » +- 0.04 and 3.88 +- 0.03. In the thorium target, 1.56 +- 0.25 fissions and 1.25 +- 0.01 atoms of /sup 233/U per incident proton were measured; the calculated values were 1.54 +- 0.01 fissions and 1.27 +- 0.01 atom/proton.« less

Seasonal uranium distributions in the coastal waters off the Amazon and Mississippi Rivers

USGS Publications Warehouse

Swarzenski, P.W.; McKee, B.A.

1998-01-01

The chemical reactivity of uranium was investigated across estuarine gradients from two of the world's largest river systems: the Amazon and Mississippi. Concentrations of dissolved (<0.45 ??m) uranium (U) were measured in surface waters of the Amazon shelf during rising (March 1990), flood (June 1990) and low (November 1991) discharge regimes. The dissolved U content was also examined in surface waters collected across estuarine gradients of the Mississippi outflow region during April 1992, August 1993, and November (1993). All water samples were analyzed for U by isotope dilution inductively coupled plasma mass spectrometry (ICP-MS). In Amazon shelf surface waters uranium increased nonconservatively from about 0.01 ??g l-1 at the river's mouth to over 3 ??g l-1 at the distal site, irrespective of river discharge stage. Observed large-scale U removal at salinities generally less than 15 implies a) that riverine dissolved U was extensively adsorbed by freshly-precipitated hydrous metal oxides (e.g., FeOOH, MnO2) as a result of flocculation and aggregation, and b) that energetic resuspension and reworking of shelf sediments and fluid muds on the Amazon shelf released a chemically reactive particle/colloid to the water column which can further scavenge dissolved U across much of the estuarine gradient. In contrast, the estuarine chemistry of U is inconclusive within surface waters of the Mississippi shelf-break region. U behavior is most likely controlled less by traditional sorption and/or desorption reactions involving metal oxides or colloids than by the river's variable discharge regime (e.g., water parcel residence time during estuarine mixing, nature of particulates, sediment storage and resuspension in the confined lower river), and plume dispersal. Mixing of the thin freshwater lens into ambient seawater is largely defined by wind-driven rather than physical processes. As a consequence, in the Mississippi outflow region uranium predominantly displays ConserVative behavior; removal is evident only during anomalous river discharge regimes. 'Products-approach' mixing experiments conducted during the Flood of 1993 suggest the importance of small particles and/or colloids in defining a depleted U versus salinity distribution.

Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lillo, Thomas; Rooyen, Isabella Van

2015-05-01

Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number ofmore » nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all grain boundaries and triple junctions contained precipitates with fission products and/or uranium, along with the differences in migration behavior between Pd, Ag and U, it was concluded that crystallographic grain boundary and triple junction parameters likely influence migration behavior.« less

Potential Treatment of Inflammatory and Proliferative Diseases by Ultra-Low Doses of Ionizing Radiations

PubMed Central

Sanders, Charles L.

2012-01-01

Ultra-low doses and dose- rates of ionizing radiation are effective in preventing disease which suggests that they also may be effective in treating disease. Limited experimental and anecdotal evidence indicates that low radiation doses from radon in mines and spas, thorium-bearing monazite sands and enhanced radioactive uranium ore obtained from a natural geological reactor may be useful in treating many inflammatory conditions and proliferative disorders, including cancer. Optimal therapeutic applications were identified via a literature survey as dose-rates ranging from 7 to 11μGy/hr or 28 to 44 times world average background rates. Rocks from an abandoned uranium mine in Utah were considered for therapeutic application and were examined by γ-ray and laser-induced breakdown fluorescence spectroscopy. The rocks showed the presence of transuranics and fission products with a γ-ray energy profile similar to aged spent uranium nuclear fuel (93% dose due to β particles and 7% due to γ rays). Mud packs of pulverized uranium ore rock dust in sealed plastic bags delivering bag surface β,γ dose-rates of 10–450 μGy/h were used with apparent success to treat several inflammatory and proliferative conditions in humans. PMID:23304108

Effect of pH on uranium(VI) biosorption and biomineralization by Saccharomyces cerevisiae.

PubMed

Zheng, X Y; Shen, Y H; Wang, X Y; Wang, T S

2018-07-01

Biosorption of radionuclides by microorganisms is a promising and effective method for the remediation of contaminated areas. pH is the most important factor during uranium biosorption by Saccharomyces cerevisiae because the pH value not only affects the biosorption rate but also affects the precipitation structure. This study investigated the effect of pH on uranium (VI) biosorption and biomineralization by S. cerevisiae. Cells have the ability to buffer the solution to neutral, allowing the biosorption system to reach an optimal level regardless of the initial pH value. This occurs because there is a release of phosphate and ammonium ions during the interaction between cells and uranium. The uranyl and phosphate ions formed nano-particles, which is chernikovite H 2 (UO 2 ) 2 (PO 4 ) 2 ·8H 2 O (PDF #08-0296), on cell surface under the initial acidic conditions. However, under the initial alkaline conditions, the uranyl, phosphate and ammonium ions formed a large amount of scale-like precipitation, which is uramphite (NH 4 )(UO 2 )PO 4 ·3H 2 O (PDF #42-0384), evenly over on cell surface. Copyright © 2018 Elsevier Ltd. All rights reserved.

Nominations for the 2017 NNSA Pollution Prevention Awards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salzman, Sonja L.; Ballesteros Rodriguez, Sonia; Lopez, Lorraine Bonds

In the field of nuclear forensics, one of the biggest challenges is to dissolve postdetonation debris for analysis. Debris generated after a nuclear detonation is a glassy material that is difficult to dissolve with chemicals. Traditionally, concentrated nitric acid, hydrofluoric acid, or sulfuric acid are employed during the dissolution. These acids, due to their corrosive nature, are not suitable for in-field/on-site sample preparations. Uranium oxides are commonly present in nuclear fuel processing plants and nuclear research facilities. In uranium oxides, the level of uranium isotope enrichment is a sensitive indicator for nuclear nonproliferation and is monitored closely by the Internationalmore » Atomic Energy Agency (IAEA) to ensure there is no misuse of nuclear material or technology for nuclear weapons. During an IAEA on-site inspection at a facility, environmental surface swipe samples are collected and transported to the IAEA headquarters or network of analytical laboratories for further processing. Uranium oxide particles collected on the swipe medium are typically dissolved with inorganic acids and are then analyzed for uranium isotopic compositions. To improve the responsiveness of on-site inspections, in-field detection techniques have been recently explored. However, in-field analysis is bottlenecked by time-consuming and hazardous dissolution procedures, as corrosive inorganic acids must be used. Corrosive chemicals are difficult to use in the field due to personnel safety considerations, and the transportation of such chemicals is highly regulated. It was therefore necessary to develop fast uranium oxide dissolution methods using less hazardous chemicals in support of the rapid infield detection of anomalies in declared nuclear processes.« less

The Proliferation Security Initiative: A Means to an End for the Operational Commander

DTIC Science & Technology

2009-05-04

The Reduced Enrichment for Research and Test Reactors ( RERTR ) Program develops technology necessary to enable the conversion of civilian...facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The RERTR Program was initiated by the U.S. Department of...processes have been developed for producing radioisotopes with LEU targets. The RERTR Program is managed by the Office of Nuclear Material Threat

Genetic and grade and tonnage models for sandstone-hosted roll-type uranium deposits, Texas Coastal Plain, USA

USGS Publications Warehouse

Hall, Susan M.; Mihalasky, Mark J.; Tureck, Kathleen; Hammarstrom, Jane M.; Hannon, Mark

2017-01-01

The coincidence of a number of geologic and climatic factors combined to create conditions favorable for the development of mineable concentrations of uranium hosted by Eocene through Pliocene sandstones in the Texas Coastal Plain. Here 254 uranium occurrences, including 169 deposits, 73 prospects, 6 showings and 4 anomalies, have been identified. About 80 million pounds of U3O8 have been produced and about 60 million pounds of identified producible U3O8 remain in place. The development of economic roll-type uranium deposits requires a source, large-scale transport of uranium in groundwater, and deposition in reducing zones within a sedimentary sequence. The weight of the evidence supports a source from thick sequences of volcanic ash and volcaniclastic sediment derived mostly from the Trans-Pecos volcanic field and Sierra Madre Occidental that lie west of the region. The thickest accumulations of source material were deposited and preserved south and west of the San Marcos arch in the Catahoula Formation. By the early Oligocene, a formerly uniformly subtropical climate along the Gulf Coast transitioned to a zoned climate in which the southwestern portion of Texas Coastal Plain was dry, and the eastern portion humid. The more arid climate in the southwestern area supported weathering of volcanic ash source rocks during pedogenesis and early diagenesis, concentration of uranium in groundwater and movement through host sediments. During the middle Tertiary Era, abundant clastic sediments were deposited in thick sequences by bed-load dominated fluvial systems in long-lived channel complexes that provided transmissive conduits favoring transport of uranium-rich groundwater. Groundwater transported uranium through permeable sandstones that were hydrologically connected with source rocks, commonly across formation boundaries driven by isostatic loading and eustatic sea level changes. Uranium roll fronts formed as a result of the interaction of uranium-rich groundwater with either (1) organic-rich debris adjacent to large long-lived fluvial channels and barrier–bar sequences or (2) extrinsic reductants entrained in formation water or discrete gas that migrated into host units via faults and along the flanks of salt domes and shale diapirs. The southwestern portion of the region, the Rio Grande embayment, contains all the necessary factors required for roll-type uranium deposits. However, the eastern portion of the region, the Houston embayment, is challenged by a humid environment and a lack of source rock and transmissive units, which may combine to preclude the deposition of economic deposits. A grade and tonnage model for the Texas Coastal Plain shows that the Texas deposits represent a lower tonnage subset of roll-type deposits that occur around the world, and required aggregation of production centers into deposits based on geologic interpretation for the purpose of conducting a quantitative mineral resource assessment.

DISPERSION HARDENING OF URANIUM METAL

DOEpatents

Arbiter, W.

1963-01-15

A method of hardening U metal involves the forming of a fine dispersion of UO/sub 2/. This method consists of first hydriding the U to form a finely divided powder and then exposing the powder to a very dilute O gas in an inert atmosphere under such pressure and temperature conditions as to cause a thin oxide film to coat each particle of the U hydride, The oxide skin prevents agglomeration of the particles as the remaining H is removed, thus preserving the small particle size. The oxide skin coatings remain as an oxide dispersion. The resulting product may be workhardened to improve its physical characteristics. (AEC)

The Past Informs the Future: An Overview of the Million Worker Study and the Mallinckrodt Chemical Works Cohort.

PubMed

Boice, John D; Ellis, Elizabeth D; Golden, Ashley P; Girardi, David J; Cohen, Sarah S; Chen, Heidi; Mumma, Michael T; Shore, Roy E; Leggett, Richard W

2018-04-01

The purpose of this paper is to present an overview of ongoing work on the Million Worker Study (MWS), highlighting some of the key methods and progress so far as exemplified by the study of workers at the Mallinckrodt Chemical Works (MCW). The MWS began nearly 25 y ago and continues in a stepwise fashion, evaluating one study cohort at a time. It includes workers from U.S. Department of Energy (DOE) Manhattan Project facilities, U.S. Nuclear Regulatory Commission (NRC) regulated nuclear power plants, industrial radiographers, U.S. Department of Defense (DoD) nuclear weapons test participants, and physicians and technologists working with medical radiation. The purpose is to fill the major gap in radiation protection and science: What is the risk when exposure is received gradually over time rather than briefly as for the atomic bomb survivors? Studies published or planned in 2018 include leukemia (and dosimetry) among atomic veterans, leukemia among nuclear power plant workers, mortality among workers at the MCW, and a comprehensive National Council on Radiation Protection and Measurements (NCRP) Report on dosimetry for the MWS. MCW has a singular place in history: the 40 tons (T) of uranium oxide produced at MCW were used by Enrico Fermi on 2 December 1942 to produce the first manmade sustained and controlled nuclear reaction, and the atomic age was born. Seventy-six years later, the authors followed the over 2,500 MCW workers for mortality and reconstructed dose from six sources of exposure: external gamma rays from the radioactive elements in pitchblende; medical x rays from occupationally required chest examinations; intakes of pitchblende (uranium, radium, and silica) measured by urine samples; radon breath analyses and dust surveys overseen by Robley Evans and Merril Eisenbud; occupational exposures received before and after employment at MCW; and cumulative radon concentrations and lung dose from the decay of radium in the work environment. The unique exposure reconstructions allow for multiple evaluations, including estimates of silica dust. The study results are relevant today. For example, NASA is interested that radium, deposited in the brain, releases high-LET alpha particles - the only human analogue, though limited, for high energy, high-Z particles (galactic cosmic rays) traveling through space that might affect astronauts on Mars missions. Don't discount the past; it's the prologue to the future!

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

PubMed

Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

2012-11-30

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste. Copyright © 2012 Elsevier B.V. All rights reserved.

Modeling of multiphase flow with solidification and chemical reaction in materials processing

NASA Astrophysics Data System (ADS)

Wei, Jiuan

Understanding of multiphase flow and related heat transfer and chemical reactions are the keys to increase the productivity and efficiency in industrial processes. The objective of this thesis is to utilize the computational approaches to investigate the multiphase flow and its application in the materials processes, especially in the following two areas: directional solidification, and pyrolysis and synthesis. In this thesis, numerical simulations will be performed for crystal growth of several III-V and II-VI compounds. The effects of Prandtl and Grashof numbers on the axial temperature profile, the solidification interface shape, and melt flow are investigated. For the material with high Prandtl and Grashof numbers, temperature field and growth interface will be significantly influenced by melt flow, resulting in the complicated temperature distribution and curved interface shape, so it will encounter tremendous difficulty using a traditional Bridgman growth system. A new design is proposed to reduce the melt convection. The geometric configuration of top cold and bottom hot in the melt will dramatically reduce the melt convection. The new design has been employed to simulate the melt flow and heat transfer in crystal growth with large Prandtl and Grashof numbers and the design parameters have been adjusted. Over 90% of commercial solar cells are made from silicon and directional solidification system is the one of the most important method to produce multi-crystalline silicon ingots due to its tolerance to feedstock impurities and lower manufacturing cost. A numerical model is developed to simulate the silicon ingot directional solidification process. Temperature distribution and solidification interface location are presented. Heat transfer and solidification analysis are performed to determine the energy efficiency of the silicon production furnace. Possible improvements are identified. The silicon growth process is controlled by adjusting heating power and moving the side insulation layer upward. It is possible to produce high quality crystal with a good combination of heating and cooling. SiC based ceramic materials fabricated by polymer pyrolysis and synthesis becomes a promising candidate for nuclear applications. To obtain high uniformity of microstructure/concentration fuel without crack at high operating temperature, it is important to understand transport phenomena in material processing at different scale levels. In our prior work, a system level model based on reactive porous media theory was developed to account for the pyrolysis process in uranium-ceramic nuclear fabrication In this thesis, a particle level mesoscopic model based on the Smoothed Particle Hydrodynamics (SPH) is developed for modeling the synthesis of filler U3O8 particles and SiC matrix. The system-level model provides the thermal boundary conditions needed in the particle level simulation. The evolution of particle concentration and structure as well as composition of composite produced will be investigated. Since the process temperature and heat flux play the important roles in material quality and uniformity, the effects of heating rate at different directions, filler particle size and distribution on uniformity and microstructure of the final product are investigated. Uncertainty issue is also discussed. For the multiphase flow with directional solidification, a system level based on FVM is established. In this model, melt convection, temperature distribution, phase change and solidification interface can be investigated. For the multiphase flow with chemical reaction, a particle level model based on SPH method is developed to describe the pyrolysis and synthesis process of uranium-ceramic nuclear fuel. Due to its mesh-free nature, SPH can easily handle the problems with multi phases and components, large deformation, chemical reactions and even solidifications. A multi-scale meso-macroscopic approach, which combine a mesoscopic model based on SPH method and macroscopic model based on FVM, FEM and FDM, can be applied to even more complicated system. In the mesoscopic model by SPH method, some fundamental mesoscopic phenomena, such as the microstructure evolution, interface morphology represented by high resolution, particle entrapment in solidification can be studied. In the macroscopic model, the heat transfer, fluid flow, species transport can be modeled, and the simulation results provided the velocity, temperature and species boundary condition necessary for the mesoscopic model. This part falls into the region of future work. (Abstract shortened by UMI.)

Gamma-ray spectroscopy measurements and simulations for uranium mining

NASA Astrophysics Data System (ADS)

Marchais, T.; Pérot, B.; Carasco, C.; Allinei, P.-G.; Chaussonnet, P.; Ma, J.-L.; Toubon, H.

2018-01-01

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration evaluation by means of gamma measurements. This paper reports gamma-ray spectra, recorded with a high-purity coaxial germanium detector, on standard cement blocks with increasing uranium content, and the corresponding MCNP simulations. The detailed MCNP model of the detector and experimental setup has been validated by calculation vs. experiment comparisons. An optimization of the detector MCNP model is presented in this paper, as well as a comparison of different nuclear data libraries to explain missing or exceeding peaks in the simulation. Energy shifts observed between the fluorescence X-rays produced by MCNP and atomic data are also investigated. The qualified numerical model will be used in further studies to develop new gamma spectroscopy approaches aiming at reducing acquisition times, especially for ore samples with low uranium content.

Quantum oscillations of nitrogen atoms in uranium nitride

NASA Astrophysics Data System (ADS)

Aczel, A. A.; Granroth, G. E.; MacDougall, G. J.; Buyers, W. J. L.; Abernathy, D. L.; Samolyuk, G. D.; Stocks, G. M.; Nagler, S. E.

2012-10-01

The vibrational excitations of crystalline solids corresponding to acoustic or optic one-phonon modes appear as sharp features in measurements such as neutron spectroscopy. In contrast, many-phonon excitations generally produce a complicated, weak and featureless response. Here we present time-of-flight neutron scattering measurements for the binary solid uranium nitride, showing well-defined, equally spaced, high-energy vibrational modes in addition to the usual phonons. The spectrum is that of a single atom, isotropic quantum harmonic oscillator and characterizes independent motions of light nitrogen atoms, each found in an octahedral cage of heavy uranium atoms. This is an unexpected and beautiful experimental realization of one of the fundamental, exactly solvable problems in quantum mechanics. There are also practical implications, as the oscillator modes must be accounted for in the design of generation IV nuclear reactors that plan to use uranium nitride as a fuel.

Quantum oscillations of nitrogen atoms in uranium nitride.

PubMed

Aczel, A A; Granroth, G E; Macdougall, G J; Buyers, W J L; Abernathy, D L; Samolyuk, G D; Stocks, G M; Nagler, S E

2012-01-01

The vibrational excitations of crystalline solids corresponding to acoustic or optic one-phonon modes appear as sharp features in measurements such as neutron spectroscopy. In contrast, many-phonon excitations generally produce a complicated, weak and featureless response. Here we present time-of-flight neutron scattering measurements for the binary solid uranium nitride, showing well-defined, equally spaced, high-energy vibrational modes in addition to the usual phonons. The spectrum is that of a single atom, isotropic quantum harmonic oscillator and characterizes independent motions of light nitrogen atoms, each found in an octahedral cage of heavy uranium atoms. This is an unexpected and beautiful experimental realization of one of the fundamental, exactly solvable problems in quantum mechanics. There are also practical implications, as the oscillator modes must be accounted for in the design of generation IV nuclear reactors that plan to use uranium nitride as a fuel.

A STUDY OF THE ACIDOSIS, BLOOD UREA, AND PLASMA CHLORIDES IN URANIUM NEPHRITIS IN THE DOG, AND OF THE PROTECTIVE ACTION OF SODIUM BICARBONATE.

PubMed

Goto, K

1917-05-01

1. The presence of an acidosis in dogs with experimental uranium nephritis is demonstrable by the Van Slyke-Stillman-Cullen method and that of Marriott. It is detected more readily by the former method. 2. This acidosis is associated with increase in the blood urea and plasma chlorides and with the appearance of albumin and casts in the urine. 3. The oral administration of sodium bicarbonate diminishes the acidosis, the increase in plasma chlorides, the amount of albumin and casts in the urine, and, to a lesser degree, the increase in the blood urea following the administration of uranium. It also diminishes the severity of the changes produced by uranium in the kidneys. 4. The oral administration of sodium bicarbonate to normal dogs raises the carbon dioxide content of the plasma as determined by the. Van Slyke-Stillman-Cullen method.

Control of a laser inertial confinement fusion-fission power plant

DOEpatents

Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

2015-10-27

A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

Estimating terrestrial uranium and thorium by antineutrino flux measurements.

PubMed

Dye, Stephen T; Guillian, Eugene H

2008-01-08

Uranium and thorium within the Earth produce a major portion of terrestrial heat along with a measurable flux of electron antineutrinos. These elements are key components in geophysical and geochemical models. Their quantity and distribution drive the dynamics, define the thermal history, and are a consequence of the differentiation of the Earth. Knowledge of uranium and thorium concentrations in geological reservoirs relies largely on geochemical model calculations. This article describes the methods and criteria to experimentally determine average concentrations of uranium and thorium in the continental crust and in the mantle by using site-specific measurements of the terrestrial antineutrino flux. Optimal, model-independent determinations involve significant exposures of antineutrino detectors remote from nuclear reactors at both a midcontinental and a midoceanic site. This would require major, new antineutrino detection projects. The results of such projects could yield a greatly improved understanding of the deep interior of the Earth.

Estimating terrestrial uranium and thorium by antineutrino flux measurements

PubMed Central

Dye, Stephen T.; Guillian, Eugene H.

2008-01-01

Uranium and thorium within the Earth produce a major portion of terrestrial heat along with a measurable flux of electron antineutrinos. These elements are key components in geophysical and geochemical models. Their quantity and distribution drive the dynamics, define the thermal history, and are a consequence of the differentiation of the Earth. Knowledge of uranium and thorium concentrations in geological reservoirs relies largely on geochemical model calculations. This article describes the methods and criteria to experimentally determine average concentrations of uranium and thorium in the continental crust and in the mantle by using site-specific measurements of the terrestrial antineutrino flux. Optimal, model-independent determinations involve significant exposures of antineutrino detectors remote from nuclear reactors at both a midcontinental and a midoceanic site. This would require major, new antineutrino detection projects. The results of such projects could yield a greatly improved understanding of the deep interior of the Earth. PMID:18172211

(Reaction mechanism studies of heavy ion induced nuclear reactions): Annual progress report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mignerey, A.C.

1988-10-01

A major experiment was performed at the Oak Ridge National Laboratory Holifield Heavy Ion Research Facility in January 1988. The primary goal of the experiment was to determine the excitation energy division in the initial stages of damped reactions. The reaction of /sup 35/Cl on /sup 209/Bi was chosen because the excited projectile-like fragments would preferentially emit light charged particles and the target-like fragments deexcite via neutron emission. This provides a means by which projectile excitations can be selected over target excitations through detection of light charged particles in coincidence with projectile-like fragments. Two experiments were performed during the pastmore » year at the Lawrence Berkeley Laboratory Bevalac in collaboration with the Wozniak-Moretto group. The first was in February 1988 and was a continuation of earlier work on La-induced reactions at intermediate energies. Beams of La with E/A = 80 and 100 MeV were used to bombard targets of C, Al, and Cu. At this time a test run was also performed using the uranium beam to see if the intensity was sufficient to use this very heavy beam for future experiments. The high intensities obtained for uranium showed that it was feasible to extend the studies of inverse reactions begun with the lanthanum beam to a heavier beam. Gold rather than uranium was chosen for our major run in August due to its low fission probability and higher beam intensity. No results are yet available for that experiment.« less

Prototype development of ion exchanging alpha detectors

NASA Astrophysics Data System (ADS)

Krupp, Dominik; Scherer, Ulrich W.

2018-07-01

In contemporary alpha particle spectrometry, the sample preparation is separated from the detection of the radionuclides. The sample preparation itself requires much time and the equipment of a radiochemistry lab. If sample preparation and detection could be combined in one step, a huge time-saving potential becomes available. One way to realize such a combination is described here. The concept was explored by simulations with the well-established computer programs SRIM and AASI. In a proof of concept, the active surface of commercially available alpha detectors was modified with sulfonic acid groups as a well-known type of cation exchanger. It was shown, that in contrast to a pristine detector, a chemically modified detector is able to extract uranium-238 and -234 selectively as uranyl cations onto the detector surface from a diluted [238/234U]uranyl acetate solution. It was possible to measure directly in the sample solution for one week or to prepare the modified detector surfaces within 30 s for measurements in conventional alpha chambers. In either case, the full width at half maximum of the measured spectra was around 100 keV, allowing a clear nuclide identification. After regenerating the cation exchanger surfaces by rinsing with hydrochloric acid the typical uranium spectra had disappeared, proving chemical bonding of the uranium. Due to the large variety of potential functional groups this new way of alpha spectrometry could be beneficial for all fields of alpha particle spectrometry, from environmental analysis, over security measurements to studies of the heaviest elements.

Depleted and natural uranium: chemistry and toxicological effects.

PubMed

Craft, Elena; Abu-Qare, Aquel; Flaherty, Meghan; Garofolo, Melissa; Rincavage, Heather; Abou-Donia, Mohamed

2004-01-01

Depleted uranium (DU) is a by-product from the chemical enrichment of naturally occurring uranium. Natural uranium is comprised of three radioactive isotopes: (238)U, (235)U, and (234)U. This enrichment process reduces the radioactivity of DU to roughly 30% of that of natural uranium. Nonmilitary uses of DU include counterweights in airplanes, shields against radiation in medical radiotherapy units and transport of radioactive isotopes. DU has also been used during wartime in heavy tank armor, armor-piercing bullets, and missiles, due to its desirable chemical properties coupled with its decreased radioactivity. DU weapons are used unreservedly by the armed forces. Chemically and toxicologically, DU behaves similarly to natural uranium metal. Although the effects of DU on human health are not easily discerned, they may be produced by both its chemical and radiological properties. DU can be toxic to many bodily systems, as presented in this review. Most importantly, normal functioning of the kidney, brain, liver, and heart can be affected by DU exposure. Numerous other systems can also be affected by DU exposure, and these are also reviewed. Despite the prevalence of DU usage in many applications, limited data exist regarding the toxicological consequences on human health. This review focuses on the chemistry, pharmacokinetics, and toxicological effects of depleted and natural uranium on several systems in the mammalian body. A section on risk assessment concludes the review.

Quantifying uranium transport rates and storage of fluvially eroded mine tailings from a historic mine site in the Grand Canyon Region

NASA Astrophysics Data System (ADS)

Skalak, K.; Benthem, A. J.; Walton-Day, K. E.; Jolly, G.

2015-12-01

The Grand Canyon region contains a large number of breccia pipes with economically viable uranium, copper, and silver concentrations. Mining in this region has occurred since the late 19th century and has produced ore and waste rock having elevated levels of uranium and other contaminants. Fluvial transport of these contaminants from mine sites is a possibility, as this arid region is susceptible to violent storms and flash flooding which might erode and mobilize ore or waste rock. In order to assess and manage the risks associated with uranium mining, it is important to understand the transport and storage rates of sediment and uranium within the ephemeral streams of this region. We are developing a 1-dimensional sediment transportation model to examine uranium transport and storage through a typical canyon system in this region. Our study site is Hack Canyon Mine, a uranium and copper mine site, which operated in the 1980's and is currently experiencing fluvial erosion of its waste rock repository. The mine is located approximately 40km upstream from the Colorado River and is in a deep, narrow canyon with a small watershed. The stream is ephemeral for the upper half of its length and sediment is primarily mobilized during flash flood events. We collected sediment samples at 110 locations longitudinally through the river system to examine the distribution of uranium in the stream. Samples were sieved to the sand size and below fraction (<2mm) and uranium was measured by gamma-ray spectroscopy. Sediment storage zones were also examined in the upper 8km of the system to determine where uranium is preferentially stored in canyon systems. This information will quantify the downstream transport of constituents associated with the Hack Canyon waste rock and contribute to understanding the risks associated with fluvial mobilization of uranium mine waste.

Depleted uranium startup of spent-fuel treatment operations at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Mariani, R.D.; Bonomo, N.L.

1995-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of Experimental Breeder Reactor II (EBR-II) spent nuclear fuel. This fuel will be treated using an electrometallurgical process in the fuel conditioning facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The process equipment is undergoing testing with depleted uranium in preparation for irradiated fuel operations during the summer of 1995.

Guinea’s 2008 Military Coup and Relations with the United States

DTIC Science & Technology

2009-07-16

Guinea’s extractive industry sector is of financial and strategic interest to the United States. In addition to gold, diamonds, uranium , and potential...agency’s legal mandate and authorities have not been clearly defined, and the CNDD has not publicly outlined how the agency is meant to interact with...Guinea’s economy relies heavily on primary commodity exports, notably bauxite (used to produce aluminum), gold, diamonds, uranium , and iron ore

Nuclear Proliferation in the Middle East: Implications for the Superpowers,

DTIC Science & Technology

1982-01-01

when Israeli nuclear scientists began extracting low grade uranium from phosphate deposits in the Negev Desert. With the encouragement of Chaim... Negev -and concomitantly assisted in de- signing the research facilities associated with the reactor. By the time the reactor went critical in 1964, a...deposits in the Negev . Since the early 1970s, an esti- mated 40-50 tons of uranium oxide has been produced annually.1 6 In addition, unconfirmed

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-12-09

Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008...gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment...technology, which it mastered by the mid-1980s. Highly-enriched uranium (HEU) is one of two types of fissile material used in nuclear weapons; the other

Argonne explains nuclear recycling in 4 minutes

ScienceCinema

Willit, Jim; Williamson, Mark; Haynes, Amber

2018-05-30

Currently, when using nuclear energy only about five percent of the uranium used in a fuel rod gets fissioned for energy; after that, the rods are taken out of the reactor and put into permanent storage. There is a way, however, to use almost all of the uranium in a fuel rod. Recycling used nuclear fuel could produce hundreds of years of energy from just the uranium we've already mined, all of it carbon-free. Problems with older technology put a halt to recycling used nuclear fuel in the United States, but new techniques developed by scientists at Argonne National Laboratory address many of those issues. For more information, visit http://www.anl.gov/energy/nuclear-energy.

From rum jungle to Wismut-reducing the environmental impact of uranium mining and milling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zuk, W.M.; Jeffree, R.A.; Levins, D.M.

1994-12-31

Australia has a long history of uranium mining. In the early days, little attention was given to environmental matters and considerable pollution occurred. Ansto has been involved in rehabilitation of a number of the early uranium mining sites, from Rum Jungle in Australia`s Northern Territory to Wismut in Germany, and is working with current producers to minimise the environmental impact of their operations. Ansto`s expertise is extensive and includes, inter alia, amelioration of acid mine drainage, radon measurement and control, treatment of mill wastes, management of tailings, monitoring of seepage plumes, mathematical modelling of pollutant transport and biological impacts inmore » a tropical environment.« less

Neutron-rich isotope production using the uranium carbide multi-foil SPES target prototype

NASA Astrophysics Data System (ADS)

Scarpa, D.; Biasetto, L.; Corradetti, S.; Manzolaro, M.; Andrighetto, A.; Carturan, S.; Prete, G.; Zanonato, P.; Stracener, D. W.

2011-03-01

In the framework of the R&D program for the SPES (Selective Production of Exotic Species) project of the Istituto Nazionale di Fisica Nucleare (INFN), production yields of neutron-rich isotopes have been measured at the Holifield Radioactive Ion Beam Facility (HRIBF, Oak Ridge National Laboratory, USA). This experiment makes use of the multi-foil SPES target prototype composed of 7 uranium carbide discs, with excess of graphite (ratio C/ U = 4 . 77 isotopes of medium mass (between 72 and 141amu), produced via proton-induced fission of uranium using a 40MeV proton beam, have been collected and analyzed for the target heated at 2000 ° C target temperature.

Fission product palladium-silicon carbide interaction in htgr fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-07-01

Interaction of fission product palladium (Pd) with the silicon carbide (SiC) layer was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors (HTGR) with an optical microscope and electron probe microanalyzers. The SiC layers were attacked locally or the reaction product formed nodules at the attack site. Although the main element concerned with the reaction was palladium, rhodium and ruthenium were also detected at the corroded areas in some particles. Palladium was detected on both the hot and cold sides of the particles, but the corroded areas and the palladium accumulations were distributed particularly on the cold side of the particles. The observed Pd-SiC reaction depths were analyzed on the assumption that the release of palladium from the fuel kernel controls the whole Pd-SiC reaction.

Biogenic non-crystalline U(IV) revealed as major component in uranium ore deposits

NASA Astrophysics Data System (ADS)

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly D.; Roebbert, Yvonne; Weyer, Stefan; Bernier-Latmani, Rizlan; Borch, Thomas

2017-06-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U(VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U(IV) generated through biologically mediated U(VI) reduction is the predominant U(IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (~58-89%) of U is bound as U(IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U(VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotope signatures, consistent with largely biotic reduction of U(VI) to U(IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U(IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.

Biogenic non-crystalline U(IV) revealed as major component in uranium ore deposits

PubMed Central

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly D.; Roebbert, Yvonne; Weyer, Stefan; Bernier-Latmani, Rizlan; Borch, Thomas

2017-01-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U(VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U(IV) generated through biologically mediated U(VI) reduction is the predominant U(IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (∼58-89%) of U is bound as U(IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U(VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotope signatures, consistent with largely biotic reduction of U(VI) to U(IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U(IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment. PMID:28569759

Biogenic non-crystalline U (IV) revealed as major component in uranium ore deposits

DOE PAGES

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly D.; ...

2017-06-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U (VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U (IV) generated through biologically mediated U (VI) reduction is the predominant U (IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (~58-89%) of U is bound as U (IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U (VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotopemore » signatures, consistent with largely biotic reduction of U (VI) to U (IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U (IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.« less

Biogenic non-crystalline U(IV) revealed as major component in uranium ore deposits.

PubMed

Bhattacharyya, Amrita; Campbell, Kate M; Kelly, Shelly D; Roebbert, Yvonne; Weyer, Stefan; Bernier-Latmani, Rizlan; Borch, Thomas

2017-06-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U (VI) ) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U (IV)  generated through biologically mediated U (VI)  reduction is the predominant U (IV)  species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (∼58-89%) of U is bound as U (IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U (VI) represent only minor components. The uranium deposit exhibited mostly 238 U-enriched isotope signatures, consistent with largely biotic reduction of U (VI) to U (IV) . This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U (IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.

Effect of mineral constituents in the bioleaching of uranium from uraniferous sedimentary rock samples, Southwestern Sinai, Egypt.

PubMed

Amin, Maisa M; Elaassy, Ibrahim E; El-Feky, Mohamed G; Sallam, Abdel Sattar M; Talaat, Mona S; Kawady, Nilly A

2014-08-01

Bioleaching, like Biotechnology uses microorganisms to extract metals from their ore materials, whereas microbial activity has an appreciable effect on the dissolution of toxic metals and radionuclides. Bioleaching of uranium was carried out with isolated fungi from uraniferous sedimentary rocks from Southwestern Sinai, Egypt. Eight fungal species were isolated from different grades of uraniferous samples. The bio-dissolution experiments showed that Aspergillus niger and Aspergillus terreus exhibited the highest leaching efficiencies of uranium from the studied samples. Through monitoring the bio-dissolution process, the uranium grade and mineralogic constituents of the ore material proved to play an important role in the bioleaching process. The tested samples asserted that the optimum conditions of uranium leaching are: 7 days incubation time, 3% pulp density, 30 °C incubation temperature and pH 3. Both fungi produced the organic acids, namely; oxalic, acetic, citric, formic, malonic, galic and ascorbic in the culture filtrate, indicating an important role in the bioleaching processes. Copyright © 2014 Elsevier Ltd. All rights reserved.

Biogenic non-crystalline U (IV) revealed as major component in uranium ore deposits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly D.

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U (VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U (IV) generated through biologically mediated U (VI) reduction is the predominant U (IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (~58-89%) of U is bound as U (IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U (VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotopemore » signatures, consistent with largely biotic reduction of U (VI) to U (IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U (IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.« less

Biogenic non-crystalline U(IV) revealed as major component in uranium ore deposits

USGS Publications Warehouse

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly; Roebbert, Yvonne; Weyer, Stefan; Bernier-Latmani, Rizlan; Borch, Thomas

2017-01-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U(VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U(IV) generated through biologically mediated U(VI) reduction is the predominant U(IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (∼58-89%) of U is bound as U(IV)to C-containing organic functional groups or inorganic carbonate, while uraninite and U(VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotope signatures, consistent with largely biotic reduction of U(VI) to U(IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U(IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.

Elemental and Isotopic Analysis of Uranium Oxide an NIST Glass Standards by FEMTOSECOND-LA-ICP-MIC-MS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ebert, Chris; Zamzow, Daniel S.; McBay, Eddie H.

2009-06-01

The objective of this work was to test and demonstrate the analytical figures of merit of a femtosecond-laser ablation (fs-LA) system coupled with an inductively coupled plasma-multi-ion collector-mass spectrometer (ICP-MIC-MS). The mobile fs-LA sampling system was designed and assembled at Ames Laboratory and shipped to Oak Ridge National Laboratory (ORNL), where it was integrated with an ICP-MIC-MS. The test period of the integrated systems was February 2-6, 2009. Spatially-resolved analysis of particulate samples is accomplished by 100-shot laser ablation using a fs-pulsewidth laser and monitoring selected isotopes in the resulting ICP-MS transient signal. The capability of performing high sensitivity, spatiallymore » resolved, isotopic analyses with high accuracy and precision and with virtually no sample preparation makes fs-LA-ICP-MIC-MS valuable for the measurement of actinide isotopes at low concentrations in very small samples for nonproliferation purposes. Femtosecond-LA has been shown to generate particles from the sample that are more representative of the bulk composition, thereby minimizing weaknesses encountered in previous work using nanosecond-LA (ns-LA). The improvement of fs- over ns-LA sampling arises from the different mechanisms for transfer of energy into the sample in these two laser pulse-length regimes. The shorter duration fs-LA pulses induce less heating and cause less damage to the sample than the longer ns pulses. This results in better stoichiometric sampling (i.e., a closer correlation between the composition of the ablated particles and that of the original solid sample), which improves accuracy for both intra- and inter-elemental analysis. The primary samples analyzed in this work are (a) solid uranium oxide powdered samples having different {sup 235}U to {sup 238}U concentration ratios, and (b) glass reference materials (NIST 610, 612, 614, and 616). Solid uranium oxide samples containing {sup 235}U in depleted, natural, and enriched abundances were analyzed as particle aggregates immobilized in a collodion substrate. The uranium oxide samples were nuclear reference materials (CRMs U0002, U005-A, 129-A, U015, U030-A, and U050) obtained from New Brunswick Laboratory-USDOE.« less

The ultimate mineral processing challenge: Recovery of rare earths, phosphorus and uranium from Florida phosphatic clay

DOE PAGES

Zhang, Patrick; Liang, Haijun; Jin, Zhen; ...

2017-11-01

We report phosphate beneficiation in Florida generates more than one tonne of phosphatic clay, or slime, per tonne of phosphate rock produced. Since the start of the practice of large-scale washing and desliming for phosphate beneficiation, more than 2 Gt of slime has accumulated, containing approximately 600 Mt of phosphate rock, 600 kt of rare earth elements (REEs) and 80 million kilograms of uranium. The recovery of these valuable elements from the phosphatic clay is one of the most challenging endeavors in mineral processing, because the clay is extremely dilute, with an average solids concentration of 3 percent, and finemore » in size, with more than 50 percent having particle size smaller than 2 μm, and it contains nearly 50 percent clay minerals as well as large amounts of magnesium, iron and aluminum. With industry support and under funding from the Critical Materials Institute, the Florida Industrial and Phosphate Research Institute in conjunction with the Oak Ridge National Laboratory undertook the task to recover phosphorus, rare earths and uranium from Florida phosphatic clay. This paper presents the results from the preliminary testing of two approaches. The first approach involves three-stage cycloning using cyclones with diameters of 12.4 cm (5 in.), 5.08 cm (2 in.) and 2.54 cm (1 in.), respectively, to remove clay minerals followed by flotation and leaching. The second approach is a two-step leaching process. In the first step, selective leaching was conducted to remove magnesium, thus allowing the production of phosphoric acid suitable for the manufacture of diammonium phosphate (DAP) in the second leaching step. The results showed that multistage cycloning with small cyclones is necessary to remove clay minerals. Finally, selective leaching at about pH 3.2 using sulfuric acid was found to be effective for removing more than 80 percent of magnesium from the feed with minimal loss of phosphorus.« less

The ultimate mineral processing challenge: Recovery of rare earths, phosphorus and uranium from Florida phosphatic clay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Patrick; Liang, Haijun; Jin, Zhen

We report phosphate beneficiation in Florida generates more than one tonne of phosphatic clay, or slime, per tonne of phosphate rock produced. Since the start of the practice of large-scale washing and desliming for phosphate beneficiation, more than 2 Gt of slime has accumulated, containing approximately 600 Mt of phosphate rock, 600 kt of rare earth elements (REEs) and 80 million kilograms of uranium. The recovery of these valuable elements from the phosphatic clay is one of the most challenging endeavors in mineral processing, because the clay is extremely dilute, with an average solids concentration of 3 percent, and finemore » in size, with more than 50 percent having particle size smaller than 2 μm, and it contains nearly 50 percent clay minerals as well as large amounts of magnesium, iron and aluminum. With industry support and under funding from the Critical Materials Institute, the Florida Industrial and Phosphate Research Institute in conjunction with the Oak Ridge National Laboratory undertook the task to recover phosphorus, rare earths and uranium from Florida phosphatic clay. This paper presents the results from the preliminary testing of two approaches. The first approach involves three-stage cycloning using cyclones with diameters of 12.4 cm (5 in.), 5.08 cm (2 in.) and 2.54 cm (1 in.), respectively, to remove clay minerals followed by flotation and leaching. The second approach is a two-step leaching process. In the first step, selective leaching was conducted to remove magnesium, thus allowing the production of phosphoric acid suitable for the manufacture of diammonium phosphate (DAP) in the second leaching step. The results showed that multistage cycloning with small cyclones is necessary to remove clay minerals. Finally, selective leaching at about pH 3.2 using sulfuric acid was found to be effective for removing more than 80 percent of magnesium from the feed with minimal loss of phosphorus.« less

High density dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

Fission fragment driven neutron source

DOEpatents

Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

1976-01-01

Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

Reduced Uranium Phases Produced from Anaerobic Reaction with Nanoscale Zerovalent Iron.

PubMed

Tsarev, Sergey; Collins, Richard N; Fahy, Adam; Waite, T David

2016-03-01

Nanoscale zerovalent iron (nZVI) has shown potential to be an effective remediation agent for uranium-contaminated subsurface environments, however, the nature of the reaction products and their formation kinetics have not been fully elucidated over a range of environmentally relevant conditions. In this study, the oxygen-free reaction of U(VI) with varying quantities of nZVI was examined at pH 7 in the presence of both calcium and carbonate using a combination of X-ray absorption spectroscopy, X-ray diffraction and transmission electron microscopy. It was observed that the structure of the reduced U solid phases was time dependent and largely influenced by the ratio of nZVI to U in the system. At the highest U:Fe molar ratio examined (1:4), nanoscale uraninite (UO2) was predominantly formed within 1 day of reaction. At lower U:Fe molar ratios (1:21), evidence was obtained for the formation of sorbed U(IV) and U(V) surface complexes which slowly transformed to UO2 nanoparticles that were stable for up to 1 year of anaerobic incubation. After 8 days of reaction at the lowest U:Fe molar ratio examined (1:110), sorbed U(IV) was still the major form of U associated with the solid phase. Regardless of the U:Fe molar ratio, the anaerobic corrosion of nZVI resulted in the slow formation of micron-sized fibrous chukanovite (Fe2(OH)2CO3) particles.

Spatial investigation of some uranium minerals using nuclear microprobe

NASA Astrophysics Data System (ADS)

Valter, Anton A.; Knight, Kim B.; Eremenko, Gelij K.; Magilin, Dmitry V.; Ponomarov, Artem A.; Pisansky, Anatoly I.; Romanenko, Alexander V.; Ponomarev, Alexander G.

2018-01-01

In this work, several individual grains of uranium minerals—uraninite with high content of Ca, Ca-rich boltwoodite, growths of uranophane with β-uranophane, and weeksite—from different uranium deposits were studied by a scanning nuclear microprobe. Particle-induced X-ray emission technique provided by the microprobe (µ-PIXE) was carried out to obtain a concentration and 2D distribution of elements in these minerals. In addition, energy dispersive X-ray spectrometry (SEM-EDS) provided by a scanning electron microscope was used. The types of minerals were determined by X-ray diffraction methods. Results of this study improved the understanding of trace elemental composition of the uranium minerals depending on their origin. Obtained signatures could be linked then to the sample provenance. Such data are important for nuclear forensics to identify the ore types and even specific ore bodies, when only small samples may be available for analysis. In this study, the µ-PIXE technique was used for obtaining the 2D distribution of trace elements that are not commonly measured by SEM-EDS at the relevant concentrations. The detected levels and precisions of elements determination by µ-PIXE were also defined. Using µ-PIXE, several micro mineral inclusions such as phosphate with high level of V and Si were identified. The age of the uranium minerals was estimated due to a significant content of radiogenic Pb that provides an additional parameter for determination of the main attributive characteristics of the minerals. This work also showed that due to its high elemental sensitivity the nuclear microprobe can be a new analytical tool for creating a nuclear forensic database from the known uranium deposits and a subsequent analysis of the intercepted illicit materials.

Spatial investigation of some uranium minerals using nuclear microprobe

NASA Astrophysics Data System (ADS)

Valter, Anton A.; Knight, Kim B.; Eremenko, Gelij K.; Magilin, Dmitry V.; Ponomarov, Artem A.; Pisansky, Anatoly I.; Romanenko, Alexander V.; Ponomarev, Alexander G.

2018-06-01

In this work, several individual grains of uranium minerals—uraninite with high content of Ca, Ca-rich boltwoodite, growths of uranophane with β-uranophane, and weeksite—from different uranium deposits were studied by a scanning nuclear microprobe. Particle-induced X-ray emission technique provided by the microprobe (µ-PIXE) was carried out to obtain a concentration and 2D distribution of elements in these minerals. In addition, energy dispersive X-ray spectrometry (SEM-EDS) provided by a scanning electron microscope was used. The types of minerals were determined by X-ray diffraction methods. Results of this study improved the understanding of trace elemental composition of the uranium minerals depending on their origin. Obtained signatures could be linked then to the sample provenance. Such data are important for nuclear forensics to identify the ore types and even specific ore bodies, when only small samples may be available for analysis. In this study, the µ-PIXE technique was used for obtaining the 2D distribution of trace elements that are not commonly measured by SEM-EDS at the relevant concentrations. The detected levels and precisions of elements determination by µ-PIXE were also defined. Using µ-PIXE, several micro mineral inclusions such as phosphate with high level of V and Si were identified. The age of the uranium minerals was estimated due to a significant content of radiogenic Pb that provides an additional parameter for determination of the main attributive characteristics of the minerals. This work also showed that due to its high elemental sensitivity the nuclear microprobe can be a new analytical tool for creating a nuclear forensic database from the known uranium deposits and a subsequent analysis of the intercepted illicit materials.

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

Development of experimental approach to examine U occurrence continuity over the extended area reconnoitory boreholes: Lostoin Block, West Khasi Hills district, Meghalaya (India).

PubMed

Kukreti, B M; Kumar, Pramod; Sharma, G K

2015-10-01

Exploratory drilling was undertaken in the Lostoin block, West Khasi Hills district of Meghalaya based on the geological extension to the major uranium deposit in the basin. Gamma ray logging of drilled boreholes shows considerable subsurface mineralization in the block. However, environmental and exploration related challenges such as climatic, logistic, limited core drilling and poor core recovery etc. in the block severely restricted the study of uranium exploration related index parameters for the block with a high degree confidence. The present study examines these exploration related challenges and develops an integrated approach using representative sampling of reconnoitory boreholes in the block. Experimental findings validate a similar geochemically coherent nature of radio elements (K, Ra and Th) in the Lostoin block uranium hosting environment with respect to the known block of Mahadek basin and uranium enrichment is confirmed by the lower U to Th correlation index (0.268) of hosting environment. A mineralized zone investigation in the block shows parent (refers to the actual parent uranium concentration at a location and not a secondary concentration such as the daughter elements which produce the signal from a total gamma ray measurement) favoring uranium mineralization. The confidence parameters generated under the present study have implications for the assessment of the inferred category of uranium ore in the block and setting up a road map for the systematic exploration of large uranium potential occurring over extended areas in the basin amid prevailing environmental and exploratory impediments. Copyright © 2015 Elsevier Ltd. All rights reserved.

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Modeling & processing of ceramic and polymer precursor ceramic matrix composite materials

NASA Astrophysics Data System (ADS)

Wang, Xiaolin

Synthesis and processing of novel materials with various advanced approaches have attracted much attention of engineers and scientists for the past thirty years. Many advanced materials display a number of exceptional properties and can be produced with different novel processing techniques. For example, AlN is a promising candidate for electronic, optical and opto-electronic applications due to its high thermal conductivity, high electrical resistivity, high acoustic wave velocity and large band gap. Large bulk AlN crystal can be produced by sublimation of AlN powder. Novel nonostructured multicomponent refractory metal-based ceramics (carbides, borides and nitrides) show a lot of exceptional mechanical, thermal and chemical properties, and can be easily produced by pyrolysis of suitable preceramic precursors mixed with metal particles. The objective of this work is to study sublimation and synthesis of AlN powder, and synthesis of SiC-based metal ceramics. For AlN sublimation crystal growth, we will focus on modeling the processes in the powder source that affect significantly the sublimation growth as a whole. To understand the powder porosity evolution and vapor transport during powder sublimation, the interplay between vapor transport and powder sublimation will be studied. A physics-based computational model will be developed considering powder sublimation and porosity evolution. Based on the proposed model, the effect of a central hole in the powder on the sublimation rate is studied and the result is compared to the case of powder without a hole. The effect of hole size on the sublimation rate will be studied. The effects of initial porosity, particle size and driving force on the sublimation rate are also studied. Moreover, the optimal growth condition for large diameter crystal quality and high growth rate will be determined. For synthesis of SiC-based metal ceramics, we will focus on developing a multi-scale process model to describe the dynamic behavior of filler particle reaction, microstructure evolution, at the microscale as well as transient fluid flow, heat transfer, and species transport at the macroscale. The model comprises of (i) a microscale model and (ii) a macroscale transport model, and aims to provide optimal conditions for the fabrication process of the ceramics. The porous media macroscale model for SiC-based metal-ceramic materials processing will be developed to understand the thermal polymer pyrolysis, chemical reaction of active fillers and transport phenomena in the porous media. The macroscale model will include heat and mass transfer, curing, pyrolysis, chemical reaction and crystallization in a mixture of preceramic polymers and submicron/nano-sized metal particles of uranium, zirconium, niobium, or hafnium. The effects of heating rate, sample size, size and volume ratio of the metal particles on the reaction rate and product uniformity will be studied. The microscale model will be developed for modeling the synthesis of SiC matrix and metal particles. The macroscale model provides thermal boundary conditions to the microscale model. The microscale model applies to repetitive units in the porous structure and describes mass transport, composition changes and motion of metal particles. The unit-cell is the representation unit of the source material, and it consists of several metal particles, SiC matrix and other components produced from the synthesis process. The reactions between different components, the microstructure evolution of the product will be considered. The effects of heating rate and metal particle size on species uniformity and microstructure are investigated.

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tufic Madi Filho; Adonis Marcelo Saliba Silva; Jose Patricio Nahuel Cardenas

2015-07-01

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to createmore » samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)« less

Measurements of natural uranium concentration and isotopic composition with permil-level precision by inductively coupled plasma-quadrupole mass spectrometry

NASA Astrophysics Data System (ADS)

Shen, Chuan-Chou; Lin, Huei-Ting; Chu, Mei-Fei; Yu, Ein-Fen; Wang, Xianfeng; Dorale, Jeffrey A.

2006-09-01

A new analytical technique using inductively coupled plasma-quadrupole mass spectrometry (ICP-QMS) has been developed that produces permil-level precision in the measurement of uranium concentration ([U]) and isotopic composition (δ234U) in natural materials. A 233U-236U double spike method was used to correct for mass fractionation during analysis. To correct for ratio drifting, samples were bracketed by uranium standard measurements. A sensitivity of 6-7 × 108 cps/ppm was generated with a sample solution uptake rate of 30 μL/min. With a measurement time of 15-20 min, standards of 30-ng uranium produced a within-run precision better than 3‰ (±2 R.S.D.) for δ234U and better than 2‰ for [U]. Replicate measurements made on standards show that a between-run reproducibility of 3.5‰ for δ234U and 2‰ for [U] can be achieved. ICP-QMS data of δ234U and [U] in seawater, coral, and speleothem materials are consistent with the data measured by other ICP-MS and TIMS techniques. Advantages of the ICP-QMS method include low cost, easy maintenance, simple instrumental operation, and few sample preparation steps. Sample size requirements are small, such as 10-14 mg of coral material. The results demonstrate that this technique can be applied to natural samples with various matrices.

Bi-Modal Model for Neutron Emissions from PuO{sub 2} and MOX Holdup

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard; Lafleur, Adrienne

2015-07-01

The measurement of uranium and plutonium holdup in plants during process activity and for decommissioning is important for nuclear safeguards and material control. The amount of plutonium and uranium holdup in glove-boxes, pipes, ducts, and other containers has been measured for several decades using both neutron and gamma-ray techniques. For the larger containers such as hot cells and glove-boxes that contain processing equipment, the gamma-ray techniques are limited by self-shielding in the sample as well as gamma absorption in the equipment and associated shielding. The neutron emission is more penetrating and has been used extensively to measure the holdup formore » the large facilities such as the MOX processing and fabrication facilities in Japan and Europe. In some case the totals neutron emission rates are used to determine the holdup mass and in other cases the coincidence rates are used such as at the PFPF MOX fabrication plant in Japan. The neutron emission from plutonium and MOX has 3 primary source terms: 1) Spontaneous fission (SF) from the plutonium isotopes, 2) The (α,n) reactions from the plutonium alpha particle emission reacting with the oxygen and other impurities, and 3) Neutron multiplication (M) in the plutonium and uranium as a result of neutrons created by the first two sources. The spontaneous fission yield per gram is independent of thickness, whereas, the above sources 2) and 3) are very dependent on the thickness of the deposit. As the effective thickness of the deposit becomes thin relative to the alpha particle range, the (α,n) reactions and neutrons from multiplication (M) approach zero. In any glove-box, there will always be two primary modes of holdup accumulation, namely direct powder contact and non-contact by air dispersal. These regimes correspond to surfaces in the glove-box that have come into direct contact with the process MOX powder versus surface areas that have not had direct contact with the powder. The air dispersal of PuO{sub 2} particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm{sup 3}, but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O{sub 2} in the PuO{sub 2} as well as any impurity contamination such as H{sub 2}O. To obtain a more quantitative estimate of the neutron (α,n) yields as a function of holdup deposit thickness, we have used MCNPX calculations to estimate the absorption of alpha particles in PuO{sub 2} holdup deposits. The powder thickness was varied from 0.1 μm to 5000 μm and the alpha particle escape probability was calculated. As would be expected, as the thickness approaches zero, the escape probability approaches 1.0, and as the thickness gets much greater than the alpha particle range (∼50 μm), the escape probability becomes small. Typically, the neutron holdup calibration measurement are performed using sealed containers of thick MOX that has all 3 sources of neutrons [SF, (α,n), and M], and no significant impurities. Thus, the calibration counting rates need to include corrections for M and (α,n) yields that are different for the holdup compared with the calibration samples. If totals neutron counting is used for the holdup measurements, the variability of the (α,n) term needs to be considered.« less

Column Testing and 1D Reactive Transport Modeling to Evaluate Uranium Plume Persistence Processes

NASA Astrophysics Data System (ADS)

Johnson, R. H.; Morrison, S.; Morris, S.; Tigar, A.; Dam, W. L.; Dayvault, J.

2015-12-01

At many U.S. Department of Energy Office of Legacy Management sites, 100 year natural flushing was selected as a remedial option for groundwater uranium plumes. However, current data indicate that natural flushing is not occurring as quickly as expected and solid-phase and aqueous uranium concentrations are persistent. At the Grand Junction, Colorado office site, column testing was completed on core collected below an area where uranium mill tailings have been removed. The total uranium concentration in this core was 13.2 mg/kg and the column was flushed with laboratory-created water with no uranium and chemistry similar to the nearby Gunnison River. The core was flushed for a total of 91 pore volumes producing a maximum effluent uranium concentration of 6,110 μg/L at 2.1 pore volumes and a minimum uranium concentration of 36.2 μg/L at the final pore volume. These results indicate complex geochemical reactions at small pore volumes and a long tailing affect at greater pore volumes. Stop flow data indicate the occurrence of non-equilibrium processes that create uranium concentration rebound. These data confirm the potential for plume persistence, which is occurring at the field scale. 1D reactive transport modeling was completed using PHREEQC (geochemical model) and calibrated to the column test data manually and using PEST (inverse modeling calibration routine). Processes of sorption, dual porosity with diffusion, mineral dissolution, dispersion, and cation exchange were evaluated separately and in combination. The calibration results indicate that sorption and dual porosity are major processes in explaining the column test data. These processes are also supported by fission track photographs that show solid-phase uranium residing in less mobile pore spaces. These procedures provide valuable information on plume persistence and secondary source processes that may be used to better inform and evaluate remedial strategies, including natural flushing.

METHOD OF PRODUCING URANIUM

DOEpatents

Foster, L.S.; Magel, T.T.

1958-05-13

A modified process is described for the production of uranium metal by means of a bomb reduction of UF/sub 4/. Difficulty is sometimes experienced in obtaining complete separation of the uranium from the slag when the process is carried out on a snnall scale, i.e., for the production of 10 grams of U or less. Complete separation may be obtained by incorporating in the reaction mixture a quantity of MnCl/sub 2/, so that this compound is reduced along with the UF/sub 4/ . As a result a U--Mn alloy is formed which has a melting point lower than that of pure U, and consequently the metal remains molten for a longer period allowing more complete separation from the slag.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

Immobilization of uranium and arsenic by injectible iron and hydrogen stimulated autotrophic sulphate reduction

NASA Astrophysics Data System (ADS)

Burghardt, D.; Simon, E.; Knöller, K.; Kassahun, A.

2007-12-01

The main object of the study was the development of a long-term efficient and inexpensive in-situ immobilization technology for uranium (U) and arsenic (As) in smaller and decentralized groundwater discharges from abandoned mining processing sites. Therefore, corrosion of grey cast iron (gcFe) and nano-scale iron particles (naFe) as well as hydrogen stimulated autotrophic sulphate reduction (aSR) were investigated. Two column experiments with sulphate reducing bacterias (SRB) (biotic gcFe , biotic naFe) and one abiotic gcFe-column experiment were performed. In the biotic naFe column, no particle translocation was observed and a temporary but intensive naFe corrosion indicated by a decrease in Eh, a pH increase and H 2 evolution. Decreasing sulphate concentrations and 34S enrichment in the column effluent indicated aSR. Fe(II) retention could be explained by siderite and consequently FeS precipitation by geochemical modeling (PhreeqC). U and As were completely immobilised within the biotic naFe column. In the biotic gcFe column, particle entrapment in open pore spaces resulted in a heterogeneous distribution of Fe-enriched zones and an increase in permeability due to preferential flow. However, Fe(II) concentrations in the effluent indicated a constant and lasting gcFe corrosion. An efficient immobilization was found for As, but not for U.

In-situ removal and characterisation of uranium-containing particles from sediments surrounding the Fukushima Daiichi Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Martin, P. G.; Griffiths, I.; Jones, C. P.; Stitt, C. A.; Davies-Milner, M.; Mosselmans, J. F. W.; Yamashiki, Y.; Richards, D. A.; Scott, T. B.

2016-03-01

Traditional methods to locate and subsequently study radioactive fallout particles have focused heavily on autoradiography coupled with in-situ analytical techniques. Presented here is the application of a Variable Pressure Scanning Electron Microscope with both backscattered electron and energy dispersive spectroscopy detectors, along with a micromanipulator setup and electron-hardening adhesive to isolate and remove individual particles before synchrotron radiation analysis. This system allows for a greater range of new and existing analytical techniques, at increased detail and speed, to be applied to the material. Using this method, it was possible to erform detailed energy dispersive spectroscopy and synchrotron radiation characterisation of material likely ejected from the Fukushima Daiichi Nuclear Power Plant found within a sediment sample collected from the edge of the 30 km exclusion zone. Particulate material sub-micron in maximum dimension examined during this work via energy dispersive spectroscopy was observed to contain uranium at levels between 19.68 and 28.35 weight percent, with the application of synchrotron radiation spectroscopy confirming its presence as a major constituent. With great effort and cost being devoted to the remediation of significant areas of eastern Japan affected by the incident, it is crucial to gain the greatest possible understanding of the nature of this contamination in order to inform the most appropriate clean-up response.

Ultrafast neutron detector

DOEpatents

Wang, Ching L.

1987-01-01

The invention comprises a neutron detector (50) of very high temporal resolution that is particularly well suited for measuring the fusion reaction neutrons produced by laser-driven inertial confinement fusion targets. The detector comprises a biased two-conductor traveling-wave transmission line (54, 56, 58, 68) having a uranium cathode (60) and a phosphor anode (62) as respective parts of the two conductors. A charge line and Auston switch assembly (70, 72, 74) launch an electric field pulse along the transmission line. Neutrons striking the uranium cathode at a location where the field pulse is passing, are enabled to strike the phosphor anode and produce light that is recorded on photographic film (64). The transmission line may be variously configured to achieve specific experimental goals.

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.

2007-03-01

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

Comparison of Grab, Air, and Surface Results for Radiation Site Characterization

NASA Astrophysics Data System (ADS)

Glassford, Eric Keith

2011-12-01

The use of proper sampling methods and sample types for evaluating sites believed to be contaminated with radioactive materials is necessary to avoid misrepresenting conditions at the site. This study was designed to investigate if the site characterization, based upon uranium contamination measured in different types of samples, is dependent upon the mass of the sample collected. A bulk sample of potentially contaminated interior dirt was collected from an abandoned metal processing mill that rolled uranium between 1948 and 1956. The original mill dates from 1910 and has a dirt floor. The bulk sample was a mixture of dirt, black and yellow particles of metal dust, and small fragments of natural debris. Small mass (approximately 0.75 grams (g)) and large mass (approximately 70g) grab samples were prepared from the bulk sample material to simulate collection of a "grab" type sample. Air sampling was performed by re-suspending a portion of the bulk sample material using a vibration table to simulate airborne contamination that might be present during site remediation. Additionally, samples of removable contaminated surface dust were collected on 47 mm diameter filter paper by wiping the surfaces of the exposure chamber used to resuspend the bulk material. Certified reference materials, one containing a precisely known quantity of U 3O8 and one containing a known quantity of natural uranium, were utilized to calibrate the gamma spectrometry measurement system. Non-destructive gamma spectrometry measurements were used to determine the content of uranium-235 (235U) at 185 keV and 143 keV, thorium-234 (234Th) at 63 keV, and protactinium-234m (234mPa) at 1001 keV in each sample. Measurement of natural uranium in small, 1 g samples is usually accomplished by radiochemical analysis in order to measure alpha particles emitted by 238U, 235U, and 234U. However, uranium in larger bulk samples can also be measured non-destructively using gamma spectrometry to detect the low energy photons from 234Th and 234mPa, the short-lived decay products of 238U, and 235U. Two sided t-tests and coefficient of variation were used to compare sampling types. The large grab samples had the lowest calculated coefficient of variation results for activity and atom percentage. The wipe samples had the highest calculated coefficient of variation of mean specific activity (dis/sec/g) for all three energies. The air filter samples had the highest coefficient of variation calculation for mean atom percentage, for both uranium isotopes examined. The data indicated that the large mass sample was the most effective at characterizing the rolling mill radioactive site conditions, since this would indicate which samples had the smallest variations compared to the mean. Additionally, measurement results of natural uranium in the samples indicate that the distribution of radioactive contamination at the sampling location is most likely non-homogeneous and that the size of the sample collected and analyzed must be sufficiently large to insure that the analytical results are truly representative of the activity present.

Combined effects of depleted uranium and ionising radiation on zebrafish embryos.

PubMed

Ng, C Y P; Pereira, S; Cheng, S H; Adam-Guillermin, C; Garnier-Laplace, J; Yu, K N

2015-11-01

In the environment, living organisms are exposed to a mixture of stressors, and the combined effects are deemed as multiple stressor effects. In the present work, the authors studied the multiple stressor effect in embryos of the zebrafish (Danio rerio) from simultaneous exposure to alpha particles and depleted uranium (DU) through quantification of apoptotic signals at 24 h post-fertilisation (hpf) revealed by vital dye acridine orange staining. In each set of experiments, dechorionated zebrafish embryos were divided into 4 groups, each having 10 embryos: Group (C) in which the embryos did not receive any further treatment; Group (IU) in which the embryos received an alpha-particle dose of 0.44 mGy at 5 hpf and were then exposed to 100 µg l(-1) of DU from 5 to 6 hpf; Group (I) in which the embryos received an alpha-particle dose of 0.44 mGy at 5 hpf and Group (U) in which the dechorionated embryos were exposed to 100 µg l(-1) of DU from 5 to 6 hpf. The authors confirmed that an alpha-particle dose of 0.44 mGy and a DU exposure for 1 h separately led to hormetic and toxic effects assessed by counting apoptotic signals, respectively, in the zebrafish. Interestingly, the combined exposure led to an effect more toxic than that caused by the DU exposure alone, so effectively DU changed the beneficial effect (hormesis) brought about by alpha-particle irradiation into an apparently toxic effect. This could be explained in terms of the promotion of early death of cells predisposed to spontaneous transformation by the small alpha-particle dose (i.e. hormetic effect) and the postponement of cell death upon DU exposure. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

A Stochastic Imaging Technique for Spatio-Spectral Characterization of Special Nuclear Material

NASA Astrophysics Data System (ADS)

Hamel, Michael C.

Radiation imaging is advantageous for detecting, locating and characterizing special nuclear material (SNM) in complex environments. A dual-particle imager (DPI) has been designed that is capable of detecting gamma-ray and neutron signatures from shielded SNM. The system combines liquid organic and NaI(Tl) scintillators to form a combined Compton and neutron scatter camera. Effective image reconstruction of detected particles is a crucial component for maximizing the performance of the system; however, a key deficiency exists in the widely used list-mode maximum-likelihood estimation-maximization (MLEM) image reconstruction technique. The steady-state solution produced by this iterative method will have poor quality compared to solutions produced with fewer iterations. A stopping condition is required to achieve a better solution but these conditions fail to achieve maximum image quality. Stochastic origin ensembles (SOE) imaging is a good candidate to address this problem as it uses Markov chain Monte Carlo to reach a stochastic steady-state solution that has image quality comparable to the best MLEM solution. The application of SOE to the DPI is presented in this work. SOE was originally applied in medical imaging applications with no mechanism to isolate spectral information based on location. This capability is critical for non-proliferation applications as complex radiation environments with multiple sources are often encountered. This dissertation extends the SOE algorithm to produce spatially dependent spectra and presents experimental result showing that the technique was effective for isolating a 4.1-kg mass of weapons grade plutonium (WGPu) when other neutron and gamma-ray sources were present. This work also demonstrates the DPI as an effective tool for localizing and characterizing highly enriched uranium (HEU). A series of experiments were performed with the DPI using a deuterium-deuterium (DD) and deuterium-tritium (DT) neutron generator, as well as AmLi, to interrogate a 13.7-kg sphere of HEU. In all cases, the neutrons and gamma rays produced from induced fission were successfully discriminated from the interrogating particles to localize the HEU. For characterization, the fast neutron and gamma-ray spectra were recorded from multiple HEU configurations with low-Z and high-Z moderation. Further characterization of the configurations used the measured neutron lifetime to show that the DPI can be used to infer multiplication.

Proliferation dangers associated with nuclear medicine: getting weapons-grade uranium out of radiopharmaceutical production.

PubMed

Williams, Bill; Ruff, Tilman A

2007-01-01

Abolishing the threat of nuclear war requires the outlawing of nuclear weapons and dismantling current nuclear weapon stockpiles, but also depends on eliminating access to fissile material (nuclear weapon fuel). The near-universal use of weapons-grade, highly enriched uranium (HEU) to produce radiopharmaceuticals is a significant proliferation hazard. Health professionals have a strategic opportunity and obligation to progress the elimination of medically-related commerce in HEU, closing one of the most vulnerable pathways to the much-feared 'terrorist bomb'.

Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.

2007-07-01

This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical densitymore » product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)« less

Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

PubMed

Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

2010-09-01

Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p < 0.05) frequently associated with lung fibrosis. Malignant tumor incidence increased linearly with dose (up to 60 Gy) with a risk of 1-1.6% Gy for MOX, similar to results for industrial plutonium oxide alone (1.9% Gy). Staining with antibodies against Surfactant Protein-C, Thyroid Transcription Factor-1, or Oct-4 showed differential labeling of tumor types. In conclusion, late effects following MOX inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

Future Scenarios for Fission Based Reactors

NASA Astrophysics Data System (ADS)

David, S.

2005-04-01

The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile inventory, capacity to be deployed, induced radiotoxicities, and R&D efforts.

Sputtering of uranium

NASA Technical Reports Server (NTRS)

Gregg, R.; Tombrello, T. A.

1978-01-01

Results are presented for an experimental study of the sputtering of U-235 atoms from foil targets by hydrogen, helium, and argon ions, which was performed by observing tracks produced in mica by fission fragments following thermal-neutron-induced fission. The technique used allowed measurements of uranium sputtering yields of less than 0.0001 atom/ion as well as yields involving the removal of less than 0.01 monolayer of the uranium target surface. The results reported include measurements of the sputtering yields for 40-120-keV protons, 40-120-keV He-4(+) ions, and 40- and 80-keV Ar-40(+) ions, the mass distribution of chunks emitted during sputtering by the protons and 80-keV Ar-40(+) ions, the total chunk yield during He-4(+) sputtering, and some limited data on molecular sputtering by H2(+) and H3(+). The angular distribution of the sputtered uranium is discussed, and the yields obtained are compared with the predictions of collision cascade theory.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no usemore » of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.« less

Method for making a uranium chloride salt product

DOEpatents

Miller, William E [Naperville, IL; Tomczuk, Zygmunt [Lockport, IL

2004-10-05

The subject apparatus provides a means to produce UCl.sub.3 in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl.sub.2 is formed. Due to is lower density, the CdCl.sub.2 rises through the Cd layer into a layer of molten LiCl--KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl.sub.2 reacts with the uranium to form UCl.sub.3 and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl.sub.3 combines with the molten salt. During production the temperature is maintained at about 600.degree. C. while after the uranium has been depleted the salt temperature is lowered, the molten salt is pressure siphoned from the vessel, and the salt product LiCl--KCl-30 mol % UCl.sub.3 is solidified.

What are the health costs of uranium mining? A case study of miners in Grants, New Mexico

PubMed Central

Jones, Benjamin A

2014-01-01

Background: Uranium mining is associated with lung cancer and other health problems among miners. Health impacts are related with miner exposure to radon gas progeny. Objectives: This study estimates the health costs of excess lung cancer mortality among uranium miners in the largest uranium-producing district in the USA, centered in Grants, New Mexico. Methods: Lung cancer mortality rates on miners were used to estimate excess mortality and years of life lost (YLL) among the miner population in Grants from 1955 to 2005. A cost analysis was performed to estimate direct (medical) and indirect (premature mortality) health costs. Results: Total health costs ranged from $2.2 million to $7.7 million per excess death. This amounts to between $22.4 million and $165.8 million in annual health costs over the 1955–1990 mining period. Annual exposure-related lung cancer mortality was estimated at 2185.4 miners per 100 000, with a range of 1419.8–2974.3 per 100 000. Conclusions: Given renewed interest in uranium worldwide, results suggest a re-evaluation of radon exposure standards and inclusion of miner long-term health into mining planning decisions. PMID:25224806

What are the health costs of uranium mining? A case study of miners in Grants, New Mexico.

PubMed

Jones, Benjamin A

2014-10-01

Uranium mining is associated with lung cancer and other health problems among miners. Health impacts are related with miner exposure to radon gas progeny. This study estimates the health costs of excess lung cancer mortality among uranium miners in the largest uranium-producing district in the USA, centered in Grants, New Mexico. Lung cancer mortality rates on miners were used to estimate excess mortality and years of life lost (YLL) among the miner population in Grants from 1955 to 2005. A cost analysis was performed to estimate direct (medical) and indirect (premature mortality) health costs. Total health costs ranged from $2·2 million to $7·7 million per excess death. This amounts to between $22·4 million and $165·8 million in annual health costs over the 1955-1990 mining period. Annual exposure-related lung cancer mortality was estimated at 2185·4 miners per 100 000, with a range of 1419·8-2974·3 per 100 000. Given renewed interest in uranium worldwide, results suggest a re-evaluation of radon exposure standards and inclusion of miner long-term health into mining planning decisions.

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

Uranium plasma radiates in the UV spectrum

NASA Technical Reports Server (NTRS)

Williams, M. D.

1973-01-01

Description of an experiment designed to produce and spectroscopically analyze a simulated gas core reactor plasma in the spectral range from 300 to 1300 A. The plasma was produced by focusing the radiation of a Q-spoiled ruby laser onto the flat surface of a pure U-238 specimen.

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Code of Federal Regulations, 2011 CFR

2011-01-01

... enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33... NUCLEAR MATERIAL Special Nuclear Material of Low Strategic Significance § 74.33 Nuclear material control... strategic significance. (a) General performance objectives. Each licensee who is authorized by this chapter...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

Statistical data on energy production and consumption and supporting information were obtained from US Bureau of Mines records supplemented by additional data obtained in Portugal. Geologic descriptions and analysis of known areas and of areas having possible future potential have been prepared by the US Geological Survey. Portugal lacks sufficient indigenous supplies of organic fuels to meet its energy demands, and so must import large quantities of petroleum and coal. Approximately 80% of Portugal's electric energy is produced by hydroelectric stations; thermal stations produce the other 20%. Portugal has produced no crude oil, natural gas, or condensate; no resources ormore » reserves in these categories are listed for Portugal in the 1976 World Energy Conference report. Until the last year or so (1980), no significant onshore petroleum exploration had been done in Portugal since 1963. Production of coal in Portugal has declined steadily to the present annual yield of about 200,000 metric tons. On the basis of estimates in only three coal fields, resources of coal of all ranks in Portugal total at least 76 million (10/sup 6/) metric tons. Uranium is mined near Viseu and Guarda in the northern part of Portugal; the Nisa mine in east-central Portugal will begin producing uranium ore in 1985 after installation of a processing plant. Portugal produced 95 metric tons of uranium oxide (U/sub 3/O/sub 8/) from ore stocks in each year from 1972 through 1974; production is assumed to have continued at the same rate since then. Geothermal energy has not been developed in mainland Portugal; however, hot springs that may have geothermal energy potential are known in the Minho district in the northwest. Geothermal energy resources exist in the Azores and a program of evaluation and exploration with technical assistance from the USGS is presently in progress there.« less

VapC toxins drive cellular dormancy under uranium stress for the extreme thermoacidophile Metallosphaera prunae.

PubMed

Mukherjee, Arpan; Wheaton, Garrett H; Counts, James A; Ijeomah, Brenda; Desai, Jigar; Kelly, Robert M

2017-07-01

When abruptly exposed to toxic levels of hexavalent uranium, the extremely thermoacidophilic archaeon Metallosphaera prunae, originally isolated from an abandoned uranium mine, ceased to grow, and concomitantly exhibited heightened levels of cytosolic ribonuclease activity that corresponded to substantial degradation of cellular RNA. The M. prunae transcriptome during 'uranium-shock' implicated VapC toxins as possible causative agents of the observed RNA degradation. Identifiable VapC toxins and PIN-domain proteins encoded in the M. prunae genome were produced and characterized, three of which (VapC4, VapC7, VapC8) substantially degraded M. prunae rRNA in vitro. RNA cleavage specificity for these VapCs mapped to motifs within M. prunae rRNA. Furthermore, based on frequency of cleavage sequences, putative target mRNAs for these VapCs were identified; these were closely associated with translation, transcription, and replication. It is interesting to note that Metallosphaera sedula, a member of the same genus and which has a nearly identical genome sequence but not isolated from a uranium-rich biotope, showed no evidence of dormancy when exposed to this metal. M. prunae utilizes VapC toxins for post-transcriptional regulation under uranium stress to enter a cellular dormant state, thereby providing an adaptive response to what would otherwise be a deleterious environmental perturbation. © 2017 Society for Applied Microbiology and John Wiley & Sons Ltd.

Speciation of uranium in surface-modified, hydrothermally treated, (UO{sub 2}){sup 2+}-exchanged smectite clays

DOE Office of Scientific and Technical Information (OSTI.GOV)

Giaquinta, D.M.; Soderholm, L.; Yuchs, S.E.

1997-08-01

A successful solution to the problem of disposal and permanent storage of water soluble radioactive species must address two issues: exclusion of the radionuclides from the environment and the prevention of leaching from the storage media into the environment. Immobilization of radionuclides in clay minerals has been studied. In addition to the use of clays as potential waste forms, information about the interactions of radionuclides with clays and how such interactions affect their speciations is crucial for successful modeling of actinide-migration. X-ray absorption spectroscopy (XAS) is used to determine the uranium speciation in exchanged and surface-modified clays. The XAS datamore » from uranyl-loaded bentonite clay are compared with those obtained after the particle surfaces have been coated with alkylsilanes. These silane films, which render the surface of the clay hydrophobic, are added in order to minimize the ability of external water to exchange with the water in the clay interlayer, thereby decreasing the release rate of the exchanged-uranium species. Mild hydrothermal conditions are used in an effort to mimic potential geologic conditions that may occur during long-term radioactive waste storage. The XAS spectra indicate that the uranyl monomer species remain unchanged in most samples, except in those samples that were both coated with an alkylsilane and hydrothermally treated. When the clay was coated with an organic film, formed by the acidic deposition of octadecyltrimethoxysilane, hydrothermal treatment results in the formation of aggregated uranium species in which the uranium is reduced from U{sup VI} to U{sup IV}.« less

Measurement of the 19F(α,n)22Na Cross Section for Nuclear Safeguards Science

NASA Astrophysics Data System (ADS)

Lowe, Marcus; Smith, M. S.; Pain, S.; Febbraro, M.; Pittman, S.; Chipps, K. A.; Thompson, S. J.; Grinder, M.; Grzywacz, R.; Smith, K.; Thornsberry, C.; Thompson, P.; Peters, W. A.; Waddell, D.; Blanchard, R.; Carls, A.; Shadrick, S.; Engelhardt, A.; Hertz-Kintish, D.; Allen, N.; Sims, H.

2015-10-01

Enriched uranium is commonly stored in fluoride matrices such as UF6. Alpha decays of uranium in UF6 will create neutrons via the 19F(α,n)22Na reaction. An improved cross section for this reaction will enable improved nondestructive assays of uranium content in storage cylinders at material enrichment facilities. To determine this reaction cross section, we have performed experiments using both forward and inverse kinematic techniques at the University of Notre Dame (forward) and Oak Ridge National Laboratory (inverse). Both experiments utilized the Versatile Array of Neutron Detectors at Low Energy (VANDLE) for neutron detection. The ORNL experiment also used a new ionization chamber for 22Na particle identification. Gating on the 22Na nuclei detected drastically reduced the background counts in the neutron time-of-flight spectra. The latest analysis and results will be presented for 19F beam energies ranging from 20-37 MeV. This work is funded in part by the DOE Office of Nuclear Physics, the National Nuclear Security Administration's Office of Defense Nuclear Nonproliferation R&D, and the NSF.

The Complex Sol-Gel Process for producing small ThO2 microspheres

NASA Astrophysics Data System (ADS)

Brykala, Marcin; Rogowski, Marcin

2016-05-01

Thorium based fuels offer several benefits compared to uranium based fuels thus they might be an attractive alternative to conventional fuel types. This study is devoted to the synthesis and the characterization of small thorium dioxide microspheres (Ø <50 μm). Their application involves using powder-free process, called the Complex Sol-Gel Process. The source sols used for the processes were prepared by the method where in the starting ascorbic acid solution the solid thorium nitrate was dissolved and partially neutralized by aqueous ammonia under pH control. The microspheres of thorium-ascorbate gel were obtained using the ICHTJ Process (INCT in English). Studies allowed to determine an optimal heat treatment with calcination temperature of 700 °C and temperature rate not higher than 2 °C/min which enabled us to obtain a crack-free surface of microspheres. The main parameters which have a strong influence on the synthesis method and features of the spherical particles of thorium dioxide are described in this article.

Behaviour and fluxes of natural radionuclides in the production process of a phosphoric acid plant.

PubMed

Bolívar, J P; Martín, J E; García-Tenorio, R; Pérez-Moreno, J P; Mas, J L

2009-02-01

In recent years there has been an increasing awareness of the occupational and public hazards of the radiological impact of non-nuclear industries which process materials containing naturally occurring radionuclides. These include the industries devoted to the production of phosphoric acid by treating sedimentary phosphate rocks enriched in radionuclides from the uranium series. With the aim of evaluating the radiological impact of a phosphoric acid factory located in the south-western Spain, the distribution and levels of radionuclides in the materials involved in its production process have been analysed. In this way, it is possible to asses the flows of radionuclides at each step and to locate those points where a possible radionuclide accumulation could be produced. A set of samples collected along the whole production process were analysed to determine their radionuclide content by both alpha-particle and gamma spectrometry techniques. The radionuclide fractionation steps and enrichment sources have been located, allowing the establishment of their mass (activity) balances per year.

Corrosion Experiments Using Spherical Uranium Powders

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powell, G. L.; Siekhaus, W. J.; Teslich, N. E.

2017-02-01

Corrosion experiments using spherical U powders are continuing with scanning electron microscopy (SEM) showing that the particles are highly textured, 5 m to 25 m diameters with 4% larger particles that are fused smaller particles. This U has a high specific surface area with no corners or back-sides, is well annealed with no machining work, and coated with a coherent oxide film, 30 nm to 300 nm thick. Exposure of this powder to low vapor pressure H 2O in the absence of O 2, i.e., a vacuum desiccator, resulted in a coherent oxide film growth of ~1 m/y, ~ 10Xmore » the growth rate in ambient air, displaying fracture along the growth plane at ~300 nm.« less

The Resin-in-pulp Process and Its Application to Ores from Brosses "BRS 10"; LE PROCEDE "RESIN IN PULP" ET SON APPLICATION AUX MINERAIS DES BROSSES "BRS 10"

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kremer, M.

1959-03-01

The resin-in-pulp process is a technical variant of the recovery process of uranium in dilute solution by means of ion exchange resins. An anion resin, XE 123, of a welldefined grain size is placed in direct contact with the pulp produced by sulfuric acid attack on ore with a low uranium content. This process is of particular value in the treatment of pulps that cannot be filtered or decanted, such as those obtained with ore from Brosses. The preparation of the pulp, the elution of the uranium, and its fixation, as well as the various factors encountered in these operations,more » are discussed. (auth)« less

Some observations on uranium carbide alloy/tungsten compatibility

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Some observations on uranium carbide alloy/tungsten compatibility.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

The health effects of depleted uranium munitions: a summary.

PubMed

2002-06-01

There has been a substantial amount of public discussion on the health effects of the use of depleted uranium (DU) munitions. In response to this concern the Royal Society set up an independent, expert working group to investigate the health effects of DU munitions. The Royal Society has now produced two reports, and this summary covering the key conclusions and recommendations from both reports. The part I report considered the increased risks of radiation-induced cancer from exposures to DU on the battlefield. Part II dealt with the risks from the chemical toxicity of uranium, non-malignant radiation effects from DU intakes, the long-term environmental consequences of the deployment of DU munitions and responses to part I including issues arising at a public meeting to discuss the part I report.

U redox fronts and kaolinisation in basement-hosted unconformity-related U ores of the Athabasca Basin (Canada): late U remobilisation by meteoric fluids

NASA Astrophysics Data System (ADS)

Mercadier, Julien; Cuney, Michel; Cathelineau, Michel; Lacorde, Mathieu

2011-02-01

Proterozoic basement-hosted unconformity-related uranium deposits of the Athabasca Basin (Saskatchewan, Canada) were affected by significant uranium redistribution along oxidation-reduction redox fronts related to cold and late meteoric fluid infiltration. These redox fronts exhibit the same mineralogical and geochemical features as the well-studied uranium roll-front deposits in siliclastic rocks. The primary hydrothermal uranium mineralisation (1.6-1.3 Ga) of basement-hosted deposits is strongly reworked to new disseminated ores comprising three distinctly coloured zones: a white-green zone corresponding to the previous clay-rich alteration halo contemporaneous with hydrothermal ores, a uranium front corresponding to the uranium deposition zone of the redox front (brownish zone, rich in goethite) and a hematite-rich red zone marking the front progression. The three zones directly reflect the mineralogical zonation related to uranium oxides (pitchblende), sulphides, iron minerals (hematite and goethite) and alumino-phosphate-sulphate (APS) minerals. The zoning can be explained by processes of dissolution-precipitation along a redox interface and was produced by the infiltration of cold (<50°C) meteoric fluids to the hydrothermally altered areas. U, Fe, Ca, Pb, S, REE, V, Y, W, Mo and Se were the main mobile elements in this process, and their distribution within the three zones was, for most of them, directly dependent on their redox potential. The elements concentrated in the redox fronts were sourced by the alteration of previously crystallised hydrothermal minerals, such as uranium oxides and light rare earth element (LREE)-rich APS. The uranium oxides from the redox front are characterised by LREE-enriched patterns, which differ from those of unconformity-related ores and clearly demonstrate their distinct conditions of formation. Uranium redox front formation is thought to be linked to fluid circulation episodes initiated during the 400-300 Ma period during uplift and erosion of the Athabasca Basin when it was near the Equator and to have been still active during the last million years. A major kaolinisation event was caused by changes in the fluid circulation regime, reworking the primary uranium redox fronts and causing the redistribution of elements originally concentrated in the uranium-enriched meteoric-related redox fronts.

Fully convolutional neural network for removing background in noisy images of uranium bearing particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tarolli, Jay G.; Naes, Benjamin E.; Butler, Lamar

A fully convolutional neural network (FCN) was developed to supersede automatic or manual thresholding algorithms used for tabulating SIMS particle search data. The FCN was designed to perform a binary classification of pixels in each image belonging to a particle or not, thereby effectively removing background signal without manually or automatically determining an intensity threshold. Using 8,000 images from 28 different particle screening analyses, the FCN was trained to accurately predict pixels belonging to a particle with near 99% accuracy. Background eliminated images were then segmented using a watershed technique in order to determine isotopic ratios of particles. A comparisonmore » of the isotopic distributions of an independent data set segmented using the neural network, compared to a commercially available automated particle measurement (APM) program developed by CAMECA, highlighted the necessity for effective background removal to ensure that resulting particle identification is not only accurate, but preserves valuable signal that could be lost due to improper segmentation. The FCN approach improves the robustness of current state-of-the-art particle searching algorithms by reducing user input biases, resulting in an improved absolute signal per particle and decreased uncertainty of the determined isotope ratios.« less