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Hot cell studies of tritium removal from and dissolution of an irradiated thoria fuel

Description: Tritium removal and dissolution studies on an irradiated thoria-based fuel were conducted in the High Level Caves at the Savannah River Laboratory (SRL). The objectives of these studies were to define the effects of key process-related parameters on tritium evolution and subsequent dissolution. The test program at SRL determined the effects on tritium removal of particle size, heating temperature, oxidation, and agitation. ThO/sub 2//UO/sub 2/ (95%/5%) fuel from the Elk River Reactor, irradiate… more
Date: January 1, 1980
Creator: Pickett, J.B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Some Scoping Experiments for a Space Reactor

Description: Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failur… more
Date: July 7, 1983
Creator: Alexander, C. A. & Ogden, J. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Voloxidation studies with UO/sub 2/ reactor fuels

Description: Voloxidation was studied experimentally in small-scale laboratory tests with irradiated UO/sub 2/ reactor fuels. The present study developed new data on the effects on the voloxidation reaction of hull length, oxygen concentration, temperature, agitation, fuel density, and fuel type and burnup. The reaction was studied by measuring weight gains, yields of U/sub 3/O/sub 8/, product densities and particle-size distributions, reaction rates, tritium release, and behavior of other off-gases (/sup 1… more
Date: January 1, 1980
Creator: Stone, J A
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Leaching studies using PNL 76-68 glass beads and UO/sub 2/ rods with Umtanum basalt and Nugget sandstone

Description: We have performed a 440-day leaching experiment, Bead Leach II, using PNL 76-68 glass beads and simulated uranium fuel rods in the presence of repository host rocks. The experiment was conducted in a single pass continuous-flow apparatus consisting of 72 channels. The experimental conditions were: 25/sup 0/C and 75/sup 0/C, flow rates of 1, 10, and 300 m1/d, and leachant solutions consisting of simulated basalt groundwater, brine, and sodium bicarbonate solution. The two host rocks studied were… more
Date: February 1, 1984
Creator: Bazan, F.; Rego, J.; Failor, R. & Coles, D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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HTGR fuels and core development program. Quarterly progress report for the period ending November 30, 1977. [Graphite and fuel irradiation; fission product release]

Description: The work reported here includes studies of basic fission product transport mechanisms, core graphite development and testing, and the development and testing of recyclable fuel systems. Materials studied include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.
Date: December 1, 1977
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Hypothetical accident conditions, free drop and thermal tests: Specification 6M

Description: The 30 gallon Specification 6M shipping container with rolled-top food pack cans as inner containers is evaluated under conditions required by 10 CFR 71.42. One kilogram of depleted uranium as UO/sub 2/ was packaged in each of the inner containers. After completion of a free drop test and a simulated thermal test, the maximum observed leakage of UO/sub 2/ for the following week was 3.2 ..mu..g. This leakage is well below the allowable leakage per week for most plutonium isotopic mixtures. Using… more
Date: May 1, 1980
Creator: Blankenship, R.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Assay of low-enriched uranium using spontaneous fission neutrons

Description: Low-enriched uranium oxide in bulk containers can be assayed for safeguards purposes, using the neutrons from spontaneous fission of /sup 238/U as a signature, to complement enrichment and mass measurement. The penetrability of the fast fission neutrons allows the inner portion of bulk samples to register. The measurement may also be useful for measuring moisture content, of significance in process control. The apparatus used can be the same as for neutron correlation counting for Pu assay. The… more
Date: January 1, 1980
Creator: Zucker, M.S. & Fainberg, A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Results from the run-beyond-cladding breach irradiation of a predefected fuel pin (RBCB-7). [LMFBR]

Description: A slit was machined through the cladding of an irradiated fuel pin and irradiation in the Experimental Breeder Reactor-II (EBR-II) was resumed. The condition of the fuel pin was continuously followed with delayed neutron (DN) monitors. When the DN signal increased to a previously established administrative limit of 800 counts per second, the test was terminated. Postirradiation examination showed the sodium-fuel reaction caused fuel pin swelling and extension of the machined slit. There was no … more
Date: February 1, 1980
Creator: Langstaff, D. C.; Almassy, M. Y. & Washburn, D. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Cs--U--O phase diagram and its application to uranium--plutonium oxide fast reactor fuel pins

Description: Portions of the cesium-uranium-oxygen system have been investigated between 873 and 1273/sup 0/K and a phase diagram has been constructed using our data and the data of other workers in the field. Thermodynamic and kinetic data have been used to examine the reactions that occur in fast-reactor fuel pins between fission-product cesium and the uranium oxide blanket. It was concluded that at the low oxygen potentials existing at the interface between the uranium-plutonium mixed-oxide and the urani… more
Date: August 1, 1977
Creator: Fee, D C; Johnson, I; Davis, S A; Shinn, W A; Staahl, G E & Johnson, C E
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Thermophysical properties of thorium and uranium systems for use in reactor safety analysis

Description: The data compilation is intended to serve as a preliminary set of thermophysical property values for use in reactor safety analyses of the Th--/sup 233/U reactor concept. The properties covered include mp, bp, enthalpy, heats of vaporization and fusion, heat capacity, thermal conductivity, density, thermal expansion, emissivity, viscosity, etc. The systems covered are Th, Th/sub 0/./sub 9/U/sub 0/./sub 1/, U, ThO/sub 2/, Th/sub 0/./sub 9/U/sub 0/./sub 1/O/sub 2/, UO/sub 2/, U/sub 0/./sub 8/Pu/s… more
Date: June 1, 1977
Creator: Fink, J.K.; Chasanov, M.G. & Leibowitz, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of molten debris freezing and wall erosion during a severe RIA test. [PWR; BWR]

Description: A one-dimensional physical model was developed to study the transient freezing of the molten debris layer (a mixture of UO/sub 2/ fuel and zircaloy cladding) produced in a severe reactivity initiated accident in-pile test and deposited on the inner surface of the test shroud wall. The wall had a finite thickness and was cooled along its outer surface by coolant bypass flow. Analyzed are the effects of debris temperature, radiation cooling at the debris layer surface, zircaloy volume ratio withi… more
Date: January 1, 1980
Creator: El-Genk, M.S. & Moore, R.L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Pellet fabrication development using thermally denitrated UO sub 2 powder

Description: Pacific Northwest Laboratory (PNL) has evaluted, on a laboratory scale, the characteristics and pellet fabrication properties of UO{sub 3} powder prepared by the thermal denitration process. Excellent quality, 96% TD (percent of theoretical density) pellets were produced from development lots of this powder. Apparently, the key to making this highly sinterable powder from uranyl nitrate is the addition of ammonium nitrate (NH{sub 4}NO{sub 3}) to the feed solution prior to thermal denitration. P… more
Date: May 1, 1992
Creator: Davis, N. C. & Griffin, C. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Consolidated Fuel-Reprocessing Program. Progress report, April 1 to June 30, 1983

Description: All research and development on fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program. Technical progress is reported in overview fashion. Conceptual studies for the proposed Breeder Reprocessing Engineering Test (BRET) have continued. Studies to date have confirmed the feasibility of modifying an existing DOE facility at Hanford, Washington. A study to measure the extent of plutonium polymerization during steam-jet transfers of nitric acid solutions… more
Date: August 1, 1983
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

Description: The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wi… more
Date: October 1, 1983
Creator: Rest, J.; Zawadski, S.A. & Piasecka, M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Failure experience in refractory alloy-clad fuel pins applicable to space nuclear power

Description: Numerous in-reactor tests were conducted in the 1960's and early 1970's to develop fuel elements for space nuclear reactors. Most of the tests emphasized refractory metal-clad UN, UC, and UO/sub 2/. Previous reviews were provided by Weaver and Scott and by Gluyas and Watson. More recently, these data were reviewed for supporting information concerning the technical feasibility issues as they relate to the current reactor designs. This paper will focus on the fuel pin failure experience to obtai… more
Date: May 1, 1984
Creator: Dutt, D.S. & Cox, C.M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program

Description: An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thic… more
Date: April 1, 1989
Creator: Senn, R.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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