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Thermal Conductivity of UO2

Description: Report discussing the thermal conductivity of uranium dioxide. The first part of the report includes the thermal effects on the substance's physical form, while the second part describes the experimental details.
Date: September 1962
Creator: Daniel, J. L. 1924-; Matolich, J., Jr. & Deem, H. W.
Partner: UNT Libraries Government Documents Department

METALLOGRAPHY OF IRRADIATED UO$sub 2$-CONTAINING FUEL ELEMENTS

Description: A description of hot laboratory metallography and techniques of operation are presented. These facilities and techniques provided a means of examining fuel elements that contain UO/sub 2/ after irradiation to high burnups. Some unusual irradiation characteristics of UO/sub 2/ were observed, and each effect is discussed. A complete explanation of the causes of such effects has not yet been obtained. (auth)
Date: January 15, 1958
Creator: Barney, W.K. & Wemple, B.D.
Partner: UNT Libraries Government Documents Department

THE REACTION OF ZIRCONIUM WITH URANIUM DIOXIDE

Description: An investigation of the causes of observed explosive reaction of zirconium-coated uranium dioxide on dissolution in nitric acid was conducted. It was concluded that such a reaction is to be expected. Possible but unconfirmed methods of alleviating the problem are suggested. (J.R.D.)
Date: June 11, 1957
Creator: Robinson, M.T.
Partner: UNT Libraries Government Documents Department

The Nitric Acid Leaching of Uranium Dioxide in a Recirculating Dissolver

Description: A brief study of the nitric acid leaching of sheared Zircaloy and stainless-steel-clad UO/sub 2/ fuels was carried out. The study was intended to demonstrate the leaching process and the use of a removable dissolver canister for transport and charge duty. In addition, the effect of sheared element length on leaching time cycles was studied. (L.T.W.)
Date: October 20, 1960
Creator: Smith, P. W.
Partner: UNT Libraries Government Documents Department

VARIABLE MODERATOR REACTOR DEVELOPMENT PROGRAM. Quarterly Progress Report No. 1

Description: Development of the boiling water UO/sub 2/ fueled Variable Moderator Reactor (VMR) is conducted under contract for the USAEC. The initiation and progress of work under Phase I of the contract, Physics and Kinetic Analysis and Initial Evaluation,'' and the preparation for Phase II, Critical Experiment and Analysis of Results,'' are reported. A hydrodynamic flow sheet representing the sequence of calculations for the BOCH program was prepared. A preliminary block diagram of the kinetics model of the VMR was prepared. Work is reported on the PUREE code which is designed to give an accurate representation of the physics of the VMR core. A fuel element fabrication speciftcation was prepared and released for quotations. A study was made to select the most appropriate material for void simulation throughout the range of interest in the VMR. (W.D.M.)
Date: August 31, 1959
Partner: UNT Libraries Government Documents Department

A process for treating uranium chips and turnings

Description: Depleted uranium (DU) chips and turnings are generated during machining of uranium metal. Because high surface area uranium is pyrophoric, the turnings are subject to spontaneous ignition in air. The oxidation of uranium to U0{sub 2} and U{sub 3}0{sub 8} is highly exothermic and therefore the reaction may be self-sustaining. A uranium fire or even rapid oxidation and thermal convection currents will cause emission of radioactive uranium oxides. In the presence of water as liquid or vapor, uranium may also oxidize into U0{sub 2} and U{sub 3}0{sub 8}-with generation of hydrogen, a flammable and explosive gas. The heat generated the water reaction may ignite the uranium or hydrogen producing a fire, explosion, or convection current resulting in some uranium oxide becoming airborne. Because the high surface area uranium has the hazardous characteristic of reactivity, it is stored immersed in diesel oil preventing contact with water or air. Los Alamos National Laboratory (LANL) has developed and constructed a process to remove the reactivity characteristic by oxidizing uranium metal to an inert product. This inert form can then be landfilled as a low-level waste. The treatment process consists of draining the packing oils, treating with sodium hypochlorite to wet-oxidize the DU to uranyl hydroxide (UO{sub 2}(OH){sub 2}), using sodium thiosulfate to reduce the (UO{sub 2}(OH)2) to U0{sub 2}, neutralizing with sodium hydroxide, and stabilizing the settled slurry in a cement matrix. The neutralized waste water is consumed at a radioactive waste water treatment facility. Studies done at LANL describe a manageable oxidation rate well within safe bounds.
Date: February 1, 1995
Creator: Dziewinska, K.; Lussiez, G. & Munger, D.
Partner: UNT Libraries Government Documents Department

DIFFUSION IN URANIUM, ITS ALLOYS, AND COMPOUNDS

Description: ABS>A review of laboratory diffusion studies on uranium and its compounds and alloys is presented. Included are results and analysis of studies on diffusion in single-phase and in multiphase U alloys, diffusion of gases in U, and diffusion in UO/sub 2/. (J.R.D.)
Date: May 1, 1961
Creator: Rothman, S.J.
Partner: UNT Libraries Government Documents Department

AVLIS modified direct denitration: UO{sub 3} powder evaluation

Description: The evaluation study demonstrated that AVLIS-enriched uranium converted to UO{sub 3} can be used to prepare UO{sub 3} pellets having densities in the range required for commercial power reactor fuel. Specifically, the program has demonstrated that MDD (Modified Direct Denitration)-derived UO{sub 2} powders can be reduced to sinterable UO{sub 2} powder using reduction techniques that allow control of the final powder characteristics; the resulting UO{sub 2} powders can be processed/sintered using standard powder preparation and pellet fabrication techniques to yield pellets with densities greater than 96% TD; pellet microstructures appear similar to those of power reactor fuel, and because of the high final pellet densities, it is expected that they would remain stable during in-reactor operation; the results of the present study confirm the results of a similar study carried out in 1982 (Davis and Griffin 1992). The laboratory processes were selected on the basis that they could be scaled up to standard commercial fuel processing. However, larger scale testing may be required to establish techniques compatible with commercial fuel fabrication techniques.
Date: February 1, 1994
Creator: Slagle, O.D.; Davis, N.C. & Parchen, L.J.
Partner: UNT Libraries Government Documents Department

Calculation of 1.25% 235U enriched UO2 solution safe slab, safe cylinder diameter, minimum safe mass, and ion exchange module for the CVDF

Description: Support calculations were performed to establish safe parameters such as fissionable material slab thickness, diameter and safe mass. These calculations were performed by MCNP for the balance of plant equipment that contains homogeneous UO{sub 2} solutions with a maximum enrichment of 1.25 Wt% {sup 235}U . The calculations were performed with the most limiting concentration of moderator and reflection so that only the safety parameters identified in the problem description need to be controlled. These calculations represent the most limiting cases for all uranium enrichments and transuranic levels due to fuel exposure for balance of plant equipment used for handling of waste water containing fissionable materials from the MCO draining and drying activities.
Date: June 26, 1997
Creator: Roblyer, S.P.
Partner: UNT Libraries Government Documents Department

Recent highlights of X-ray magnetic scattering studies from surfaces

Description: In this work, recent studies of surface magnetism, as observed by x-ray scattering techniques, are described. The experiments were concerned with uranium dioxide crystals for which x-ray resonance effects enhance the small magnetic signal from the surface. The main result is that, in contrast to the bulk which exhibits a discontinuous magnetic ordering transition, both the (001) and (111) surface layers order continuously. This is reminiscent of the general phenomenon of surface wetting, but had not been previously observed for magnetic materials. Magnetic reflectivity studies show further that the near-surface magnetic layers are more disordered than layers deep in the bulk, even at low temperatures.
Date: December 31, 1998
Creator: Watson, G.M.; Gibbs, D. & Lander, G.H.
Partner: UNT Libraries Government Documents Department

EXTRUDED CERAMIC NUCLEAR FUEL DEVELOPMENT PROGRAM. Final Report

Description: Urania rods 6-in. long and 0.475-in. in diameter were extruded and sintered to densities exceeding 94% of the theoratical urania density. The rods dropped freely through a straight metal tube 8-in. long with an internal diameter 0.004-in. greater than the diameter of the rods. All properties of the extruded and sintered rods relevant to their use as a nuclear fuel material were at least equal to the corresponding properties of pressed and sintered urania pellets. Extruded and sintered urania rods can be produced with standard ceramic-industry machinery. From preliminary estimates it appears that extrusions may be produced more cheaply than pellets. (auth)
Date: January 23, 1961
Partner: UNT Libraries Government Documents Department

THE MEASUREMENT OF OXYGEN TO METAL RATIO IN SOLID SOLUTIONS OF URANIUM AND PLUTONIUM DIOXIDES

Description: A survey was made of methods potentially useful for the determination of the oxygen to metal ratio in mixed oxides of uranium and plutonium. A gravimetric method was selected as being the most promising for adaptation in a short period of time. Development of the technique resulted in a reliable method which meets the requirements for unirradiated mixed oxide fuel samples. The method, based upon an equilibrium weight at 700 deg C in dry hydrogen, was shown to be capable of measurement of O/(Pu + U) ratios in 20% PuO/sub 2/--80% UO/sub 2/ pellets with a standard deviation of plus or minus 0.001. (auth)
Date: May 31, 1963
Creator: Lyon, W.L.
Partner: UNT Libraries Government Documents Department

Explosive Interaction of Molten UO2 and Liquid Sodium

Description: The interim report presented describes a continuation of the work reported in ANL-7890, Interaction of Sodium with Molten Uranium dioxide and Stainless Steel Using a Dropping Mode of Contact. In the current study, sodium was injected into a pool of molten uranium dioxide. The experiment consistently produced vapor explosions, both with the injection nozzle above and beneath the surface of the uranium dioxide. Although the efficiency of the conversion of thermal to mechanical energy was small (due in part to very conservative data analysis and an inefficient geometry), the results did demonstrate that there is no intrinsic reason why reactor materials cannot produce a vapor explosion.
Date: March 1976
Creator: Armstrong, D. R.; Goldfuss, G. T. & Gebner, R. H.
Partner: UNT Libraries Government Documents Department

Transient Energy Transfer by Conduction and Radiation for a Sudden Contact Between Molten UO₂ and Sodium

Description: The transient energy transfer following a sudden contact between molten uranium dioxide and sodium has been investigated, taking into consideration both conduction and internal thermal radiation in uranium dioxide. Analytical expressions for the contact-interface temperature valid for small times are derived. Illustrative calculations indicate that on a time scale relevant to fuel-coolant interactions, internal radiation of molten uranium dioxide should have an insignificant effect on the contact-interface temperature between molten uranium dioxide and sodium. It thus appears that for the purpose of assessing the potential for an explosive fuel-coolant interaction, the contact-interface temperature may be adequately determined based on consideration of pure conduction.
Date: 1978?
Creator: Cho, D. H. & Chan, S. H.
Partner: UNT Libraries Government Documents Department

Cracking and Healing Behavior of UO2 as Related to Pellet-Cladding Mechanical Interaction : Interim Report, July 1976

Description: A direct-electrical-heating apparatus has been designed and fabricated to investigate those nuclear-fuel-related phenomena involved in the gap closure-bridging annulus formation mechanism that can be reproduced in an out-of-reactor environment. Prototypic light-water-reactor uranium dioxide fuel-pellet temperature profiles have been generated utilizing high flow rates (approximately 700 liters/min) of helium coolant gas, and a re-circulating system has been fabricated to permit tests of up to 1000 h. Simulated light-water-reactor single- and multiple-thermal-cycle experiments will be conducted on both unclad and ceramic (fused silica) clad uranium dioxide pellet stacks. A laser dilatometer is used to measure pellet dimensional increase continuously during thermal cycling. Acoustic emissions from thermal-gradient cracking have been detected and correlated with crack length and crack area. The acoustic emissions are monitored continuously to provide instantaneous information about thermal-gradient cracking. Post-test metallography and fracture-mechanics measurements are utilized to characterize cracking and crack healing.
Date: October 1976
Creator: Kennedy, C. R.; Yaggee, F. L.; Voglewede, J. C.; Kupperman, D. S.; Wrona, B. J.; Ellingson, W. A. et al.
Partner: UNT Libraries Government Documents Department

Test E3 on High-Energy Transient Meltdown of Irradiated UO₂ in a TREAT Mark-II Loop

Description: Three ''gassy'' irradiated uranium dioxide pins were run past failure in a 35-ms natural TREAT transient in TREAT loop Test E3 to guide in choosing between two modeling assumptions regarding energy conversion from sodium vaporization in an irradiated-oxide-fuel/coolant interaction (FCI): (1) The trapped fission gas enhances fuel fragmentation upon failure, accelerates fuel-coolant mixing, and thus promotes an energetic FCI, or, conversely, (2) the trapped fission gas blankets the fuel and reduces the rate of heat transfer to sodium, preventing an energetic FCI.
Date: December 1977
Creator: Deitrich, L. W.; Dickerman, C. E.; Willis, F. L.; Purviance, R. T.; Schmidt, K. J.; Agrawal, A. K. et al.
Partner: UNT Libraries Government Documents Department

Intergranular fracture in UO2: derivation of traction-separation law from atomistic simulations

Description: In this study, the intergranular fracture behavior of UO2 was studied by molecular dynamics simulations using the Basak potential. In addition, the constitutive traction-separation law was derived from atomistic data using the cohesive-zone model. In the simulations a bicrystal model with the (100) symmetric tilt E5 grain boundaries was utilized. Uniaxial tension along the grain boundary normal was applied to simulate Mode-I fracture. The fracture was observed to propagate along the grain boundary by micro-pore nucleation and coalescence, giving an overall intergranular fracture behavior. Phase transformations from the Fluorite to the Rutile and Scrutinyite phases were identified at the propagating crack tips. These new phases are metastable and they transformed back to the Fluorite phase at the wake of crack tips as the local stress concentration was relieved by complete cracking. Such transient behavior observed at atomistic scale was found to substantially increase the energy release rate for fracture. Insertion of Xe gas into the initial notch showed minor effect on the overall fracture behavior.
Date: October 1, 2013
Creator: Zhang, Yongfeng; Millett, Paul C; Tonks, Michael R; Bai, Xian-Ming & Biner, S Bulent
Partner: UNT Libraries Government Documents Department