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CRITICAL EXPERIMENTS FOR THE PRELIMINARY DESIGN OF THE ARGONNE HIGH FLUX REACTOR

Description: Critical experiments were performed with two assemblies simulating a cold clean, and an end-of-cycle,- Argonne High Flux Reactor, core. Data were obtained for flux distributions; cadmium ratios; temperature and void coefficients; and control rod, beam hole, and reflector worths. The data obtained furnished confirmation of theoretical predictions. The peak 2200-m/sec flux per unit power was measured as 3 x 10/sup 7/ n/(cm/sup 2/)(sec)(watt) for both cores. The two cores had internal H/sub 2/O th… more
Date: June 1, 1961
Creator: de Villiers, J.W.L. ed.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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FEASIBILITY OF Pu$sup 239$-U$sup 235$-FUELED CORES TO PREDICT Pu$sup 239$- FUELED CORE DIMENSIONS

Description: Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region … more
Date: June 1, 1962
Creator: Meneghetti, D. & Ishikawa, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Preliminary Results of High-Temperature Bare U$sup 235$-C Critical Assembly Measurements

Description: The influence of temperature on the critical buckling or bare graphite assemblies with various carbon-to-uranium235 molar ratios has been measured. A range from l185: 1 to 2l,690: 1 was covered, for 45 to 1205'F. Preliminary results indicate that the fractional rate of change of critical buckling with core temperature varies monotonically with C/U2as ratio by a factor of five over the factor-of-eighteen range in gross C/U2as ratio. This quantity appears to approach asymptotically a value n… more
Date: June 1, 1961
Creator: Finke, R. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Surface density representation of critical configurations of U(5)O/sub 2/ planar and linear arrays

Description: A criticality data base for planar and linear arrays of low enriched damp uranium dioxide powder contained in cylindrical units is generated with the Monte Carlo criticality program KENO-IV. This information is correlated with the Limiting Surface Density model, sigma(m) = c/sub 2/(m/sub 0/-m), refining the bulky KENO data base into a simple table of correlation constants c/sub 2/ and m/sub 0/. A solid angle representation of the KENO data base is provided. The solid angles were analytically de… more
Date: June 1, 1980
Creator: Maudlin, P.J. & Thomas, J.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Comparison of measured and calculated uranium isotopic concentrations in cascade streams at the Paducah Gaseous Diffusion Plant

Description: A test has been performed at the Paducah Gaseous Diffusion Plant (PGDP) in connection with studies for the US Arms Control and Disarmament Agency on the possibility of utilizing measurements of the concentrations of the minor uranium isotopes in /sup 235/U enrichment cascade external streams as a safeguards technique (MIST). This is the fourth plant test that has been performed in connection with the MIST studies, the first three having been done at the Oak Ridge Gaseous Diffusion Plant (ORGDP)… more
Date: June 16, 1982
Creator: Blumkin, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Application of adjusted data in calculating fission-product decay energies and spectra. [ADENA code]

Description: The code ADENA, which approximately calculates fission-product beta and gamma decay energies and spectra in 19 or fewer energy groups from a mixture of /sup 235/U and /sup 239/Pu fuels, is described. The calculation uses aggregate, adjusted data derived from a combination of several experiments and summation results based on the ENDF/B-V fission-product file. The method used to obtain these adjusted data and the method used by ADENA to calculate fission-product decay energy with an absorption c… more
Date: June 1, 1982
Creator: George, D. C.; LaBauve, R. J. & England, T. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Integrated system for production of neutronics and photonics calculational constants. Volume 15, Part D. The LLL Evaluated Nuclear Data Library (ENDL): descriptions of individual evaluations for Z = 90 to 98

Description: Evaluation procedures used to produce sets of evaluated data for the 33 heavy isotopes that fall in the range Z = 90 to Z = 98 are described. At the beginning of the discussion for each individual isotope, a computer-generated listing is given which summarizes the main properties of the data sets that are contained in the evaluation. (RWR)
Date: June 17, 1977
Creator: Howerton, R.J. & MacGregor, M.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Experimental comparison of the active well coincidence counter with the random driver

Description: A direct comparison has been made between the IAEA Active Well Coincidence Counter (AWCC) and the CMB-8 Random Driver. The comparison included an experimental evaluation of precision, counting rate, accuracy, penetrability, stability, and the effect of sample inhomogeneity. Samples used in the evaluation included highly enriched U/sub 3/O/sub 8/, U/sub 3/O/sub 8/ mixed withe qgraphite, highly enriched uranium metal discs, and depleted uranium metal. These materials are typical of the samples of… more
Date: June 1, 1979
Creator: Menlove, H. O.; Ensslin, N. & Sampson, T. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Considerations for an active and passive scanner to assay nuclear waste drums

Description: Radioactive wastes are generated at many DOE laboratories, military facilities, fuel fabrication and enrichment plants, reactors, hospitals, and university research facilities. At all of these sites, wastes must be separated, packaged, categorized, and packed into some sort of container--usually 208-L (55-gal) drums--for shipment to waste-storage sites. Prior to shipment, the containers must be labeled, assayed, and certified; the assay value determines the ultimate disposition of the waste con… more
Date: June 8, 1990
Creator: Martz, H. E.; Azevedo, S. G.; Roberson, G. P.; Schneberk, D. J.; Koenig, Z. M. & Camp, D.C. (Lawrence Livermore National Lab., CA (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Atlas of photoneutron cross sections obtained with monoenergetic photons. [Between 10 and 50 MeV]

Description: The data included in this supplement replace earlier preliminary results presented in the Bicentennial Edition of the Atlas of Photoneutron Cross Sections (UCRL-78482, 1976). The nuclei included are /sup 13/C, /sup 18/O, /sup 55/Mn, /sup 59/Co, /sup 186/ /sup 188/ /sup 189/ /sup 192/Os, /sup 232/Th, and /sup 235/ /sup 236/ /sup 238/U, all from measurements performed at Lawrence Livermore Laboratory.
Date: June 1, 1979
Creator: Berman, B.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Gamma-ray spectrometric determination of UF/sub 6/ assay with 1 percent precision for international safeguards. Part 1: product and feed in 1S and 2S sample cylinders

Description: The method is based on counting the 186-keV gamma rays emitted by /sup 235/U using a Pb-collimated Ge(Li) detector. Measurements of fifty UF/sub 6/ product and feed cylinders reveal the following precisions and counting times: Product - 2S, 0.98% (600 s); Feed - 2S, 0.48% (2500 s); Product - 1S, 0.62% (1000 s); Feed - 1S, 0.73% (3000 s). A 1% precision is desired for variables - attributes verification measurements of /sup 235/U assay in UF/sub 6/ sample cylinders for safeguards inspections by … more
Date: June 15, 1981
Creator: Ricci, E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Beta and gamma decay heat measurements between 0.1s-50,000s for neutron fission of {sup 235}U, {sup 238}U and {sup 239}Pu. Progress report, March 1, 1993--January 31, 1994

Description: Decay heat measurements following the fast fission of {sup 238}U are well underway. The He-jet system and spectrometers were moved to the 1 MW research reactor to gain sufficient fast neutron flux for these measurements. On the Van de Graaff accelerator, the He-jet capillary has been shortened so that beta and gamma measurements following the thermal neutron fission of {sup 235}U could be made down to delay times near 0.1 s. Gamma-ray response functions are now well characterized for gamma ener… more
Date: June 1, 1994
Creator: Schier, W. A. & Couchell, G. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Characterizing and improving passive-active shufflers for assays of 208-Liter waste drums

Description: A passive and active neutron shuffler for 208-L waste drums has been used to perform over 1500 active and 500 passive measurements on uranium and plutonium samples in 28 different matrices. The shuffler is now better characterized and improvements have been implemented or suggested. An improved correction for the effects of the matrix material was devised from flux-monitor responses. The most important cause of inaccuracies in assays is a localized instead of a uniform distribution of fissile m… more
Date: June 1, 1992
Creator: Rinard, P. M.; Adams, E. L.; Menlove, H. O. & Sprinkle, J. K. Jr.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A new fuel loading design for the Advanced Neutron Source

Description: A new fuel loading design has been developed for the Advanced Neutron Source Reactor. In this reactor the combination of a small core volume and high power results in a very high power density. Using a direct optimization procedure the thermal-hydraulic margins for oxide temperature drop, centerline temperature and incipient boiling (and thus critical heat flux) were maximized to increase the limiting thermal power from 298 MW to 346 MW compared to the previous fuel grading, while maintaining t… more
Date: June 1, 1994
Creator: Gehin, J. C.; Renier, J. P. & Worley, B. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Status of U-235 fission as a cross section standard. [10 keV to 20 MeV]

Description: A review is made of the current status of /sup 235/U fission cross section data from thermal to 20 MeV neutron energies. The accuracy achieved is compared with the 1 percent accuracy required of a reaction-cross-section standard throughout this range. The energy ranges from thermal to 10 keV, 10 keV to 0.8 MeV and 0.8 to 20 MeV are considered separately because of the different experimental techniques required in each. The goal of normalizing all fission cross sections to the thermal value and … more
Date: June 25, 1976
Creator: Carlson, G. W. & Czirr, J. B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Disposal of defense spent fuel and HLW at the Idaho Chemical Processing Plant

Description: Irradiated nuclear fuel has been reprocessed at the Idaho Chemical Processing Plant (ICPP) since 1953 to recover uranium-235 and krypton-85 for the US Department of Energy (DOE). The resulting acidic high-level radioactive waste (HLW) has been solidified to a calcine since 1963 and stored in stainless steel underground bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage at the ICPP. In April, 1992, DOE announced that spent fuel… more
Date: June 1, 1993
Creator: Ermold, L. F.; Loo, H. H.; Klingler, R. D.; Herzog, J. D. & Knecht, D. A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Initial Report for the Aquifer Background Study: Summary of Uranium and Plutonium Data from INEEL Groundwater Samples

Description: As part of the “Aquifer Background Study,” Los Alamos National Laboratory (LANL) under contract with the Idaho National Engineering and Environmental Laboratory (INEEL) has undertaken a study to determine uranium and plutonium abundances and isotopic composition in groundwater samples collected at the INEEL. To date, four samples have been analyzed for uranium and plutonium and an additional nine samples have been analyzed for uranium. It is expected that several more samples will be analyzed f… more
Date: June 1, 2003
Creator: Roback, Robert C. & Koeppen, Don L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Traditional and Model Based Assay of Irregular Geometry Items

Description: The Analytical Development Section (ADS) of SRNL was requested to perform a waste disposal assay of two heater boxes which had been used in the HB Line dissolvers. They had been sent to SRNL for study to make recommendations on how to prevent future failure of the units when they were replaced. The study having been completed, the units needed to be characterized prior to sending to Solid Waste for disposal. An assay station consisting of a turntable, HPGe detector, CANBERRA Inspector, transmis… more
Date: June 15, 2005
Creator: MOORE, FRANK S. & SALAYMEH, SALEEM
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Calculation of resonance self shielding in $sup 235$U

Description: Analytic and Monte Carlo calculations of the Bramblett- Czirr /sup 235/U transmission experiment STANuc!. Sci. Eng. 35, 350-357 (1969)! were performed. The numbcr of fissions from the uncollided flux could be calculated as a function of depth. The results of experiment and Monte Carlo calculations are tabulated and plotted; fission and total cross sections for /sup 235/U are shown. In view of the agreement between the analytic and Monte Carlo calculations, it appeared that the ENDF/B-III cross … more
Date: June 28, 1973
Creator: Plechaty, E. F. & Cullen, D. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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MANUAL FOR THE PREPARATION OF SIMULATED FUEL REPROCESSING WASTE SOLUTION

Description: Preparation of synthetic wastes from the processing of aluminum-clad natural uranium, aluminum-enriched uranium alloy, zirconium-uranium, stainless steel-UO/sub 2/, and stainless steel-UO/sub 2/-MgO fuel elements is described. These preparations will be used in waste disposal studies. (auth)
Date: June 20, 1958
Creator: Lacy, W.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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