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Experiences with fast breeder reactor education in laboratory and short course settings

Description: The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented.
Date: January 17, 1983
Creator: Waltar, A.E.
Partner: UNT Libraries Government Documents Department

Integral data for fast reactors

Description: Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections.
Date: January 1, 1988
Creator: Collins, P. J.; Poenitz, W. P. & McFarlane, H. F.
Partner: UNT Libraries Government Documents Department

System design description of forced-convection molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4

Description: Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF/sub 2/-ThF/sub 4/-UF/sub 4/ fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705/sup 0/C (1050 to 1300/sup 0/F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed.
Date: November 1, 1976
Creator: Huntley, W. R. & Silverman, M. D.
Partner: UNT Libraries Government Documents Department

Time-series investigation of anomalous thermocouple responses in a liquid-metal-cooled reactor

Description: A study was undertaken using SAS software to investigate the origin of anomalous temperature measurements recorded by thermocouples (TCs) in an instrumented fuel assembly in a liquid-metal-cooled nuclear reactor. SAS macros that implement univariate and bivariate spectral decomposition techniques were employed to analyze data recorded during a series of experiments conducted at full reactor power. For each experiment, data from physical sensors in the tests assembly were digitized at a sampling rate of 2/s and recorded on magnetic tapes for subsequent interactive processing with CMS SAS. Results from spectral and cross-correlation analyses led to the identification of a flow rate-dependent electromotive force (EMF) phenomenon as the origin of the anomalous TC readings. Knowledge of the physical mechanism responsible for the discrepant TC signals enabled us to device and justify a simple correction factor to be applied to future readings.
Date: March 24, 1988
Creator: Gross, K.C.; Planchon, H.P. & Poloncsik, J.
Partner: UNT Libraries Government Documents Department

Physics and feasibility study of the Fast-Mixed Spectrum Reactor concept

Description: Reactor physics and fuel cycle studies, coordinated with heat transfer and material science and structural analysis work has indicated the feasibility potential of the coupled Fast-Mixed Spectrum Reactor (FMSR) concept. This concept employs what are considered reasonable extrapolations of present fast breeder reactor technology to achieve a once-through-and-store reactor fuel cycle. Since the fuel cycle for this reactor is intended to use only natural or depleted uranium for its equilibrium feed, the resultant reactor would have excellent anti-proliferation characteristics. It would also extend utilization of natural uranium resources by a factor of about 15 relative to LWR reactors when on its equilibrium fuel cycle; startup requirements would of course reduce this factor.
Date: January 1, 1979
Creator: Fischer, G.J.; Kouts, H.J.C.; Cerbone, R.J.; Shenoy, S.; Durston, C.; Ludewig, H. et al.
Partner: UNT Libraries Government Documents Department

Gas core reactors for actinide transmutation and breeder applications. Annual report

Description: This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.
Date: April 1, 1978
Creator: Clement, J.D. & Rust, J.H.
Partner: UNT Libraries Government Documents Department

Cost-competitive, inherently safe LMFBR pool plant

Description: The Cost-Competitive, Inherently Safe LMFBR Pool Plant design was prepared in GFY 1983 as a joint effort by Rockwell International and the Argonne National Laboratory with major contributions from the Bechtel Group, Inc.; Combustion Engineering, Inc.; the Chicago Bridge and Iron Company; and the General Electric Company. Using current LMFBR technology, many innovative features were developed and incorporated into the design to meet the ultimate objectives of the Breeder Program, i.e., energy costs competitive with LWRs and inherent safety features to maintain the plant in a safe condition following assumed accidents without requiring operator action. This paper provides a description of the principal features that were incorporated into the design to achieve low cost and inherent safety.
Date: January 1, 1984
Creator: McDonald, J.S.; Brunings, J.E.; Chang, Y.I.; Seidensticker, R.W. & Hren, R.R.
Partner: UNT Libraries Government Documents Department

Summary of the hydraulic evaluation of LWBR (LWBR development program)

Description: The principal hydraulic performance features of the Light Water Breeder Reactor are summarized in this report. The calculational models and procedures used for prediction of reactor flow and pressure distributions under steady-state and transient operating conditions are described. Likewise, the analysis models for evaluation of the static and dynamic performance characteristics of the hydraulically-balanced and hydraulically-buffered movable-fuel reactivity-control system are outlined. An extensive test program was conducted for qualification of the subject LWBR hydraulic evaluation models. The projected LWBR hydraulic performance is shown to fulfill design objectives and functional requirements.
Date: April 1, 1981
Creator: Stout, J.W.; Lerner, S.; McWilliams, K.D. & Turner, J.R. (eds.)
Partner: UNT Libraries Government Documents Department

Liquid Metal Fast Breeder Reactors: a bibliography

Description: This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.
Date: November 1, 1980
Creator: Raleigh, H.D. (ed.)
Partner: UNT Libraries Government Documents Department

Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

Description: The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department

Liquid Metal Fast Breeder Reactors: a bibliography

Description: This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.
Date: November 1, 1980
Creator: Raleigh, H.D. (ed.)
Partner: UNT Libraries Government Documents Department

Method for reliability analysis of complex reactor systems. [LMFBR]

Description: A method and a computer code for efficient and accurate reliability analyses of complex reactor systems are described and illustrated through an example. The method permits realistic analyses through its ability to accurately model and evaluate instantaneous and average unavailabilities for large systems with dependencies. The component models can include continuously monitored, non-repairable, and periodically tested components which are subject to failures resulting from components which are subject to failures resulting from component demands, stand-by conditions, human errors associated with testing and repair, as well as failures during actual operation. The numerical process used is efficient and allows analysis of general system configurations with arbitrary scheduling of maintenance operations.
Date: January 1, 1982
Creator: Elerath, J.G.; Vaurio, J.K. & Wood, A.P.
Partner: UNT Libraries Government Documents Department

Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

Description: Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values.
Date: May 1, 1980
Creator: Williams, C.L. & Beus, S.G.
Partner: UNT Libraries Government Documents Department

Benefits of vertical and horizontal seismic isolation for LMR (liquid metal reactor) nuclear reactor units

Description: Seismic isolation has been shown to be able to reduce transmitted seismic force and lower response accelerations of a structure. When applied to nuclear reactors, it will minimize seismic influence on the reactor design and provide a design which is less site dependent. In liquid metal reactors where components are virtually at atmospheric pressure but under severe thermal conditions, thin-walled structures are generally used for primary systems. Thin-walled structures, however, have little inherent seismic resistance. The concept of seismic isolation therefore offers a viable and effective approach that permits the reactor structures to better withstand thermal and seismic loadings simultaneously. The majority of published work on seismic isolation deals with use of horizontal isolation system only. In this investigation, however, local vertical isolation is also provided for the primary system. Such local vertical isolation is found to result in significant benefits for major massive components, such as the reactor cover, designed to withstand vertical motions and loadings. Preliminary estimations on commodity savings of the primary system show that, with additional local vertical isolation, the savings could be twice that estimated for horizontal isolation only. The degree of effectiveness of vertical isolation depends on the diameter of the reactor vessel. As the reactor vessel diameter increases, the vertical seismic effects become more pronounced and vertical isolation can make a significant contribution.
Date: January 1, 1988
Creator: Wu, Ting-shu; Chang, Y.W. & Seidensticker, R.W.
Partner: UNT Libraries Government Documents Department

DOS: the discrete-ordinates system. [LMFBR]

Description: The Discrete Ordinates System determines the flux of neutrons or photons due either to fixed sources specified by the user or to sources generated by particle interaction with the problem materials. It also determines numerous secondary results which depend upon flux. Criticality searches can be performed. Numerous input, output, and file manipulation facilities are provided. The DOS driver program reads the problem specification from an input file and calls various program modules into execution as specified by the input file.
Date: September 1, 1982
Creator: Rhoades, W. A. & Emmett, M. B.
Partner: UNT Libraries Government Documents Department

Successive collision calculation of resonance absorption (AWBA Development Program)

Description: The successive collision method for calculating resonance absorption solves numerically the neutron slowing down problem in reactor lattices. A discrete energy mesh is used with cross sections taken from a Monte Carlo library. The major physical approximations used are isotropic scattering in both the laboratory and center-of-mass systems. This procedure is intended for day-to-day analysis calculations and has been incorporated into the current version of MUFT. The calculational model used for the analysis of the nuclear performance of LWBR includes this resonance absorption procedure. Test comparisons of results with RCPO1 give very good agreement.
Date: July 1, 1980
Creator: Schmidt, E. & Eisenhart, L.D.
Partner: UNT Libraries Government Documents Department

Structural materials for breeder reactor cores and coolant circuits

Description: The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these components.
Date: February 1, 1984
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Growth rates of breeder reactor fuel. Final report

Description: During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus.
Date: January 1, 1979
Creator: Ott, K O
Partner: UNT Libraries Government Documents Department

Potential of large heterogeneous reactors. [LMFBR]

Description: Different 1200 MWe heterogeneous core configurations have been evaluated and compared with a homogeneous 1200 MWe LMFBR design. This study discusses the neutronic coupling concept which was employed to discriminate between different heterogeneous core designs. Reactivity control, breeding performance, safety parameters, fuel pin optimization, transient behavior and thermal performance were investigated in detail and results of these analyses are presented.
Date: January 1, 1977
Creator: Barthold, W.P.; Tzanos, C.P. & Beitel, J.C.
Partner: UNT Libraries Government Documents Department

Systematic approach for constructing low sodium void heterogeneous cores. [LMFBR]

Description: In unprotected loss-of-flow transients in large LMFBRs sodium voiding may result in high ramp rates of positive reactivity addition leading to rates of energy release that may exceed the design limits of the containment. Therefore, there is an incentive in designing reactors of low sodium void reactivity to reduce the positive reactivity ramp rates arising from sodium voiding. A systematic way of constructing heterogeneous configurations that have a near zero value of sodium void reactivity is presented.
Date: January 1, 1977
Creator: Tzanos, C.P. & Barthold, W.P.
Partner: UNT Libraries Government Documents Department

ac power control in the Core Flow Test Loop

Description: This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report.
Date: January 1, 1980
Creator: McDonald, D.W.
Partner: UNT Libraries Government Documents Department

Analysis of sodium valve reliability data at CREDO. [LMFBR]

Description: The Centralized Reliability Data Organization (CREDO) has been established at Oak Ridge National Laboratory (ORNL) by the Department of Energy to provide a centralized source of data for reliability/maintainabilty analysis of advanced reactor systems. The current schedule calls for develoment of the data system at a moderate pace, with the first major distribution of data in late FY-1980. Continuous long-term collection of engineering, operating, and event data has been initiated at EBR-II and FFTF.
Date: January 1, 1979
Creator: Bott, T F & Haas, P M
Partner: UNT Libraries Government Documents Department

Overview of the ORNL THORS program

Description: The ORNL Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility is an engineering-scale sodium loop (40 l/s, 2 MW) in which electrically-heated bundles made to simulate segments of LMFBR core assemblies are subjected to thermal-hydraulic testing under a wide range of flows and power levels. For the past ten years the THORS Program has provided the US LMFBR effort with thermal-hydraulic results for nominal operating conditions, with inlet and heated-zone blockages, and under flow-power mismatch conditions sufficient to produce sodium boiling. THORS objectives, past accomplishments and testing history are reviewed.
Date: January 1, 1981
Creator: Wantland, J.L.
Partner: UNT Libraries Government Documents Department