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Design and Economic Evaluation of Mobile Blankets for Fast Reactors

Description: Report evaluating the design characteristics and limitations of mobile blankets for breeder reactors. This also includes economic considerations for each tested blanket. Appendices begin on page 40.
Date: March 10, 1964
Creator: Klickman, A. E.; Ball, G. L.; Edwards, J. J.; Jens, W. H.; Segal, B. M.; Amorosi, A. et al.
Partner: UNT Libraries Government Documents Department

Carbide Fuels in Fast Reactors

Description: Abstract: Cladding and fuel material processing prospects are reviewed, and fuel system possibilities for near-term (~1 mill/kwh) and long-range (<0.5 mil/kmh) fuel cycles are described.
Date: September 15, 1965
Creator: Wheelock, C. W.
Partner: UNT Libraries Government Documents Department

Statistical Identification of Effective Input Variables

Description: A statistical sensitivity analysis procedure has been developed for ranking the input data of large computer codes in the order of sensitivity-importance. The method is economical for large codes with many input variables, since it uses a relatively small number of computer runs. No prior judgmental elimination of input variables is needed. The screening method is based on stage-wise correlation and extensive regression analysis of output values calculated with selected input value combinations. The regression process deals with multivariate nonlinear functions, and statistical tests are also available for identifying input variables that contribute to threshold effects, i.e., discontinuities in the output variables. A computer code SCREEN has been developed for implementing the screening techniques. The efficiency has been demonstrated by several examples and applied to a fast reactor safety analysis code (Venus-II). However, the methods and the coding are general and not limited to such applications.
Date: September 1982
Creator: Vaurio, J. K.
Partner: UNT Libraries Government Documents Department

Fast Reactor Core Design Parameter Study

Description: Report describing parametric studies of eleven fast reactor fuel systems undertaken to determine the design and economic factors for producing electricity. The methods used for making the parametric studies are described, as well as the results of these studies. Appendices begin on page 103.
Date: March 1960
Creator: Atomic Power Development Associates
Partner: UNT Libraries Government Documents Department

Reactor Physics Studies in the Engineering Mockup Critical Assembly of the Fast Test Reactor

Description: Reactor physics studies in the Engineering Mockup Critical (EMC) assembly of the Fast Test Reactor (FTR) facility are reported. The study included measurements of the neutron spectrum, Doppler effect, sodium-void worth, reaction rates, subassembly worths, material replacement worths, and FTR control, safety and shim rod worths. Each of these physics studies were made in a clean plutonium (low-Pu-240) fuel composition environment and a dirty plutonium (high-Pu-240) fuel composition environment. The fuel studies were in support of determining the attendant effects of operation and safety of utilizing Light Water Reactor (LWR) plutonium fuel in the FTR. Comparison of the measured and calculated results are presented.
Date: July 1976
Creator: Pond, R. B.
Partner: UNT Libraries Government Documents Department

POLYFAIL: A Program for Identification of Multiple Fuel Failures with Gas Tagging

Description: This report describes the development of the computer code POLYFAIL for identification of fuel failures in fast reactors or light-water reactors that use gas tagging. POLYFAIL implements a sophisticated numerical algorithm known as the method of barycentric coordinates. The code can treat problems involving up to four simultaneous tag releases in a tagging system characterized by three independent tag ratios. The sensitivity of the multiple-failure-resolution technique has been optimized by incorporation of a newly developed ratio weighting scheme. Several example problems are provided to demonstrate operation of the code under single-leaker and various postulated multiple-leaker situations.
Date: 1981
Creator: Gross, Kenny C.
Partner: UNT Libraries Government Documents Department

Development of a MK-II Loop to Simulate Reactor Hydraulic Conditions

Description: The Mk-IIC Integral Loop was modified to provide an in-pile experimental apparatus that would simulate the subassembly coolant flow rate and inlet pressure head of the Fast Test Reactor (FTR). There were two main design changes. First, the safety dump tanks were removed from the Mk-IIC loop and replaced by a second annular linear induction pump (ALIP). Second, a flow restricting orifice was sized so that the hydraulic requirements of prototypical test-section coolant velocity and pressure head would be achieved. The resulting redesigned loop was used for the in-pile TREAT transient over-power Test H6, which investigated fuel sweep-out and coolability following fuel-pin failure under hydraulic conditions typical of the FTR. The procedure reported here will help in the design of advanced TREAT vehicles such as the Mk-III loop.
Date: January 1979
Creator: Page, R. J. & Robinson, L. E.
Partner: UNT Libraries Government Documents Department

Fluctuation Analysis of Fast Reactor Safety Experiments in TREAT

Description: Statistical fluctuations of measured signals about their mean are related to physical processes in fuel-failure experiments. Signal variance, correlation, and spectral density are shown to be sensitive measures to instrument response characteristics, of flow-blockage formation, and of boiling phenomena. This sensitivity is demonstrated by a series of examples that use test data from the E6, E7, and L5 experiments. A mathematical model of the Mark-II loop is developed to predict both the mean and the fluctuation behavior of measured test parameters. The analysis is extended to include signal forecasting by the ARIMA time-series model. Techniques that are used to identify the model and to estimate the model parameters are discussed in detail. It is shown that departure of real-time data from the on-line forecasts is a powerful tool for the rapid detection of off-normal conditions. A description of the experiments and the data-reduction process is given in the Appendices.
Date: November 1978
Creator: Doerner, R. C.; Meek, C. C.; Hurt, R. F. & Pekarsky, M. I.
Partner: UNT Libraries Government Documents Department

A Comparison of Long-Lived, Prolieration Resistant Fast Reactors

Description: Nuclear power is expected to play a significant role in meeting future electricity needs, and in significantly reducing emissions compared to fossil-fueled power plants. However, the next generation of nuclear power plants will be expected to demonstrate significant advancements in economics, safety, waste disposal, and proliferation resistance. Many reactor types have been proposed for “Generation IV”, some of which have been fast reactors. The work discussed in here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal of the entire project is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The goal of the work presented in this paper is to investigate and compare a variety of possible fuel types, looking for optimum economics for an actinide burning, low cost of electricity, reactor design using sodium or lead-bismuth as the coolant.
Date: September 1, 2001
Creator: Weaver, Kevan Dean; Herring, James Stephen & Mac Donald, Philip Elsworth
Partner: UNT Libraries Government Documents Department

A Comparison of Long-Lived, Proliferation Resistant Fast Reactors

Description: Various methods have been proposed to transmute and thus consume the current inventory of trans-uranic waste that exists in spent light-water-reactor fuel. These methods include both critical and sub-critical systems. The neutronics of metallic and nitride fuels loaded with 20-30wt% light-water-reactor plutonium plus minor actinides for use in a lead-bismuth and sodium cooled fast reactor are discussed, with an emphasis on the fuel cycle life and isotopic content. Calculations show that core life can extend beyond 20 years, and the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from 0.5 to 0.9 g/MWd.
Date: September 1, 2001
Creator: Weaver, Kevan Dean; Herring, James Stephen & Mac Donald, Philip Elsworth
Partner: UNT Libraries Government Documents Department

Effect of Temperature and Reactivity Changes in Operation of the Los Alamos Plutonium Reactor

Description: The operation of the Fast Reactor is considered in terms of normal equilibrium conditions and normal shut-down. The proposed loading, control rod adjustment and subsequent "floating" operation are discussed. Safety devices and interlocks are described. Temperature and reactivity changes are examined with respect to various system failures, phase changes, and "flashing" of the reactor. Slow changes due to faulty slug cooling are also considered. The calculations were initially based upon 10 kw operation. Performance tests of the mercury system now indicate that 20-kw operation may be feasible.
Date: May 28, 1948
Creator: Hall, David B. & Hall, Jane H.
Partner: UNT Libraries Government Documents Department

Physics of Reactor Safety, Quarterly Report: October-December 1975

Description: Quarterly progress report summarizing work done in Argonne National Laboratory's Applied Physics Division including: reactor safety research and technical coordination of the RSR safety analysis program by members of the Reactor Safety Appraisals Group, Monte Carlo analysis of safety-related critical assembly experiments by members of the Theoretical Fast Reactor Physics Group, and planning of DEMO safety-related critical experiments by members of the Zero Power Reactor (ZPR) Planning and Experiments Group.
Date: 197u
Creator: Argonne National Laboratory. Applied Physics Division.
Partner: UNT Libraries Government Documents Department

Computer-Code Formulation for Three-Dimensional HEXCAN Response Coupled with Internal Hydrodynamics

Description: A procedure is described for the dynamic analysis of a fast-reactor hexagonal subassembly. The internals of the fuel subassembly are treated by an axisymmetric hydrodynamic code REXCO-HT which, among other properties, possesses a model of an MFCI pressure source. The housing of the fuel subassembly is handled by the SADCAT code, which is based on a triangular finite element in three-dimensional space. The code is used to illustrate the discrepancies involved if the hexcan is modeled by a cylinder of the same thickness. A study is also made of the reduction of cylinder thickness such that the same final cylindrical deformation can be predicted. A discussion in arriving at such an equivalence is offered.
Date: March 1976
Creator: Marchertas, A. H. & Julke, R. T.
Partner: UNT Libraries Government Documents Department

Two-Facility Approach to Investigations of Fast Reactor Meltdowns

Description: Report issued by the APDA over studies conducted on hypothetical fast reactor meltdown accidents. As stated in the summary, "the purpose of this study is to investigate the technical feasibility of conducting fast reactor meltdown experiments to determine a more realistic maximum accident and to provide a better evaluation of fast reactor safety" (p. 7). This report includes tables, and illustrations.
Date: January 15, 1963
Creator: Doyle, T. A.; Page, E. M.; Nicholson, R. B.; Hagstrom, J. T. & Nims, J. B.
Partner: UNT Libraries Government Documents Department

Physics of Reactor Safety, Quarterly Report: October-December 1977

Description: Quarterly progress report summarizing work done in Argonne National Laboratory's Applied Physics Division and Components Technology Division. The work in the Applied Physics Division includes reports on reactor safety program by members of the Reactor Safety Appraisals Group, Monte Carlo analysis of safety-related critical assembly experiments by members of the Theoretical Fast Reactor Physics Group, and Planning of Safety-Related (ZPR) Planning and Experiments Group. Work on reactor core thermal-hydraulic code development performed in the Components Technology Division is also included in this report.
Date: March 1978
Creator: Argonne National Laboratory. Applied Physics Division.
Partner: UNT Libraries Government Documents Department