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Engineering-Scale Liquid Cadmium Cathode Experiments

Description: Recovery of transuranic actinides (TRU) using electrorefining is a process being investigated as part of the Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). TRU recovery via electrorefining onto a solid cathode is very difficult as the thermodynamic properties of transuranics are not favourable for them to remain in the metal phase while significant quantities of uranium trichloride exist in the electrolyte. Theoretically, the concentration of transuranics in the electrolyte must be approximately 106 greater than the uranium concentration in the electrolyte to produce a transuranic deposit on a solid cathode. Using liquid cadmium as a cathode contained within a LiCl-KCl eutectic salt, the co-deposition of uranium and transuranics is feasible because the activity of the transuranics in liquid cadmium is very small. Depositing transuranics and uranium in a liquid cadmium cathode (LCC) theoretically requires the concentration of transuranics to be two to three times the uranium concentration in the electrolyte. Three LCC experiments were performed in an Engineering scale elecdtrorefiner, which is located in the argon hot cell of the Fuel Conditioning Facility at the Materials and Fuels Complex on the Idaho National Laboratory. Figure 1 contains photographs of the LCC assembly in the hot cell prior to the experiment and a cadmium ingot produced after the first LCC test. Figure 1. Liquid Cadmium Cathode (left) and Cadmium Ingot (right) The primary goal of the engineering-scale liquid cadmium cathode experiments was to electrochemically collect kilogram quantities of uranium and plutonium via a LCC. The secondary goal was to examine fission product contaminations in the materials collected by the LCC. Each LCC experiment used chopped spent nuclear fuel from the blanket region of the Experimental Breeder Reactor II loaded into steel baskets as the anode with the LCC containing 26 kg of cadmium metal. In each experiment, between one ...
Date: August 1, 2006
Creator: Vaden, D; Westphal, B. R.; Li, S. X.; Johnson, T. A.; Davies, K. B. & Pace, D. M.
Partner: UNT Libraries Government Documents Department

Actinide recycle in LMFBRs as a waste management alternative

Description: A strategy of actinide burnup in fast reactor systems has been investigated as an approach for reducing the long term hazards and storage requirements of the actinide waste elements and their decay daughters. The actinide recycle studies also included plutonium burnup studies in the event that plutonium is no longer required as a fuel. Particular emphasis was placed upon the timing of the recycle program, the requirements for separability of the waste materials, and the impact of the actinides on the reactor operations and performance. It is concluded that actinide recycle and plutonium burnout are attractive alternative waste management concepts. 25 refs., 14 figs., 34 tabs.
Date: August 21, 1979
Creator: Beaman, S.L.
Partner: UNT Libraries Government Documents Department

High-temperature flaw assessment procedure

Description: Described is the background work performed jointly by the Electric Power Research Institute in the United States, the Central Research Institute of Electric Power Industry in Japan and Nuclear Electric plc in the United Kingdom with the purpose of developing a high-temperature flaw assessment procedure for reactor components. Existing creep-fatigue crack-growth models are reviewed, and the most promising methods are identified. Sources of material data are outlined, and results of the fundamental deformation and crack-growth tests are discussed. Results of subcritical crack-growth exploratory tests, creep-fatigue crack-growth tests under repeated thermal transient conditions, and exploratory failure tests are presented and contrasted with the analytical modeling. Crack-growth assessment methods are presented and applied to a typical liquid-metal reactor component. The research activities presented herein served as a foundation for the Flaw Assessment Guide for High-Temperature Reactor Components Subjected to Creep-Fatigue Loading published separately. 30 refs., 108 figs., 13 tabs.
Date: August 1, 1991
Creator: Ruggles, M.B. (Oak Ridge National Lab., TN (United States)); Takahashi, Y. (Central Research Inst. of Electric Power Industry, Tokyo (Japan)) & Ainsworth, R.A. (Nuclear Electric plc, Barnwood (UK))
Partner: UNT Libraries Government Documents Department

Reference fuel studies. Seventh quarterly report May-July 1976. [LMFBR]

Description: Task 3 of Contract AT03-76SF78003 consists of the following programs: fuel rod chemistry and thermodynamics; fuel rod engineering; fuel irradiations testing and analysis; reference structural materials. The four parts are closely interrelated and in combination are aimed at providing a sound basis for the design and performance evaluation of LMFBR mixed oxide fuel rods.
Date: August 1, 1976
Partner: UNT Libraries Government Documents Department

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

Description: This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.
Date: August 1, 1984
Creator: Todreas, N.E.; Cheng, S.K. & Basehore, K.
Partner: UNT Libraries Government Documents Department

Added mass and damping coefficients for hexagonal tube arrays

Description: An analytical investigation of the fluid coupling effects from an array of hexagonal cylindrical ducts undergoing harmonic oscillations is presented. A closed form solution for the velocity and pressure is obtained under a thin gap approximation for the case of moderate frequencies. From this solution, the usual viscous and inertial fluid coupling coefficients are easily obtained. These analytically derived coefficients indicate a strong dependence upon gap spacing and oscillating Reynolds number.
Date: August 1, 1979
Creator: Wilson, D.E.
Partner: UNT Libraries Government Documents Department

The effect of product form upon fatigue-crack growth behavior in Alloy 718: Additional results

Description: A previous study had characterized the fatigue-crack growth behavior of four wrought product forms (sheet, plate, bar and forging) from a single heat of Alloy 718 and concluded that there were no consistent trends in the crack growth rate results that could be attributed to product form variability. The present study adds one additional product form (gas-tungsten-arc weldments) from the same heat, and compares the behavior to that exhibited by the wrought product forms. Two different precipitation heat-treatments were employed at each of five test temperatures. 11 refs., 5 figs., 3 tabs.
Date: August 1, 1980
Creator: James, L.A.
Partner: UNT Libraries Government Documents Department

Leakage and motion detection system for the flexible-joint assembly, large-scale LMFBR

Description: Flexible joint assemblies in the primary sodium piping to the large scale LMFBR pressure vessel are designed to accommodate thermal expansion of the piping system. To monitor the performance of the flexible joint assemblies during reactor operation, sodium leakage and flexible joint motion detection/sensors are specified. This report describes the mechanical/hydraulic portions only of the leak detection and motion monitoring system. A tentative choice of displacement transducer and its readout equipment is presented. However, the required EI and C portion of the monitoring system is not covered.
Date: August 1, 1978
Creator: Gaspar, N.L.
Partner: UNT Libraries Government Documents Department

AI-MSG modification work plan. [LMFBR]

Description: This document contains the Work Plan for the modification of the AI Steam Generator for tests in Large Leak Test Rig. This Work Plan describes the objectives, scope of work, schedule and manpower, end items, and meetings and reports required for the modification.
Date: August 20, 1973
Creator: Page, J.P.
Partner: UNT Libraries Government Documents Department

Effect of simulated thermal shield motion on nuclear instrument response: measurements and calculations (LWBR Development Program)

Description: An experiment has been performed to determine the effect of motion of a thermal shield on the neutron signal expected from ex-core detectors. Using a mockup of the LWBR reactor vessel, thermal shield, and core barrel in conjunction with a /sup 252/Cf neutron source, the change in detector signal with displacement of the various components was investigated. It was found that moving the thermal shield would produce a significant change in detector signal, although the effect was smaller than would be produced by moving the source and core barrel together. The results were substantiated by two-dimensional discrete-ordinate calculations.
Date: August 1, 1979
Creator: Schick, W.C. Jr.; Emert, C.J.; Shure, K. & Natelson, M.
Partner: UNT Libraries Government Documents Department

Condensation of fuel onto the above-core structure during an LMFBR core-disruptive accident

Description: Condensation of a pure, saturated vapor onto a vertical, melting substrate is analyzed for both one- and two-material situations. Examination of the one-material situation indicates that the solution to the full transient condensation-induced melting problem may be approximated by using a transient, conduction-only model for short times and a steady-state, flowing-film model for long times. This concept is extrapolated to the two-material situation in order to obtain a simulation of the transient solution. The models are applied to the specific case of uranium dioxide condensing onto solid stainless steel. Condensate solidification occurs for this pair of materials; however, this solidification may be neglected without introducing a serious error in the other phase-change rates. The condensation heat flux for this pair of materials is a very weak function of the initial substrate temperature and the vapor temperature. The results of this analysis have applications in the area of LMFBR accident analysis.
Date: August 1, 1977
Creator: Erdman, C.A. Reynolds, A.B.
Partner: UNT Libraries Government Documents Department

Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

Description: Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.
Date: August 1, 1985
Creator: Clayton, J.C.
Partner: UNT Libraries Government Documents Department

Elevated-Temperature Mechanical Properties of an Advanced Type 316 Stainless Steel

Description: Type 316FR stainless steel is a candidate material for the Japanese Demonstration Fast Breeder Reactor Plant to be built in Japan early in the next century. Like type 316L(N), it is a low-carbon grade of stainless steel with a more closely specified nitrogen content and chemistry optimized to enhance elevated-temperature performance. Early in 1994, under sponsorship of The Japan Atomic Power Company, work was initiated at Oak Ridge National Laboratory (ORNL) aimed at obtaining an elevated-temperature mechanical-properties database on a single heat of this material. The product form was 50-mm plate manufactured by the Nippon Steel Corporation. Data include results from long-term creep-rupture tests conducted at temperatures of 500 to 600 C with test times up to nearly 40,000 h, continuous-cycle strain-controlled fatigue test results over the same temperature range, limited creep-fatigue data at 550 and 600 C, and tensile test properties from room temperature to 650 C. The ORNL data were compared with data obtained from several different heats and product forms of this material obtained at Japanese laboratories. The data were also compared with results from predictive equations developed for this material and with data available for type 316 and type 316L(N) stainless steel.
Date: August 1, 1999
Creator: Brinkman, C.R.
Partner: UNT Libraries Government Documents Department