Sample records for fissium

  1. In-core measurements of U-5 wt % fissium alloy thermal conductivity

    SciTech Connect

    Betten, P.R.


    Instrumented subassemblies have inserted into the Experimental Breeder Reactor II (EBR-II) in order to monitor thermal-hydraulic phenomena. For one such subassembly, a time history of the in-core themocouples was evaluated and used to determine the fuel thermal conductivity. Although several researchers have evaluated fuel conductivity for unirradiated conditions, little data is available for long term irradiation. Further, most of the data has been evaluated under laboratory conditions which, while providing exact measurements, may be missing important facets of in-core behavior. The purpose of this paper is to present the in-situ measurements of thermal conductivity over the subassembly lifetime. 7 refs.



    Zegler, S.T.


    The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

  3. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    NASA Astrophysics Data System (ADS)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun


    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  4. Potentiometric determination of uranium in organic extracts

    SciTech Connect

    Bodnar, L.Z.


    The potentimetric determination of uranium in organic extracts was studied. A mixture of 30% TBP, (tributylphosphate), in carbon tetrachloride was used, with the NBL (New Brunswick Laboratory) titrimetric procedure. Results include a comparative analysis performed on organic extracts of fissium alloys vs those performed on aqueous samples of the same alloys which had been treated to remove interfering elements. Also comparative analyses were performed on sample solutions from a typical scrap recovery operation common in the uranium processing industry. A limited number of residue type materials, calciner products, and presscakes were subjected to analysis by organic extraction. The uranium extraction was not hindered by 30% TBP/CCl/sub 4/. To fully demonstrate the capabilities of the extraction technique and its compatibility with the NBL potentiometric uranium determination, a series of uranium standards was subjected to uranium extraction with 30% TBP/CCl/sub 4/. The uranium was then stripped out of the organic phase with 40 mL of H/sub 3/PO/sub 4/, 15 mL of H/sub 2/0, and 1 mL of 1M FeSO/sub 4/ solution. The uranium was then determined in the aqueous phosphoric phase by the regular NBL potentiometric method, omitting only the addition of another 40 mL of H/sub 3/PO/sub 4/. Uranium determinations ranging from approximately 20 to 150 mg of U were successfully made with the same accuracy and precision normally achieved. 8 tables. (DP)

  5. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R


    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  6. Power coupling in TREAT M-Series: New experimental results from M7CAL and updated analyses

    SciTech Connect

    Robinson, W R; Bauer, T H


    Experiments and methods used to determine power coupling of test fuel to the TREAT reactor during six recent metal-fueled sodium loop tests (M2-M7) are described. Previously reported calibration work on a three-pin test configuration with uranium-fissium fuel is updated (M2CAL). Additional results on a two-pin test configuration with the Integral Fast Reactor (IFR) reference fuel (uranium-zirconium and uranium-plutonium-zirconium) are reported (M7CAL). The peak axial low-level, steady-state (LLSS) fresh fuel pin power coupling factors for the IFR fuel compositions were determined from radiochemical analysis of fuel segments. A large data base of uranium-zirconium neutron flux monitor wire measurements were compiled to extend the fuel measurements to high-power transient conditions by comparing the measured power couplings from high and low-power wire irradiations. Power coupling results were obtained in both a full-slotted and a half-slotted TREAT core configuration. Relative power coupling measurements are compared to calculations for the three different types of fuel; U/Fs, U/Zr and U/Pu/Zr. Estimates of power coupling including corrections accounting for the effect on the power coupling of isotopic depletion and fuel swelling as the fuel undergoes burnup are presented for planning and analysis of tests M5, M6 and M7.