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Spallator - accelerator breeder

Description: The concept involves the use of spallation neutrons produced by interaction of a high energy proton (1 to 2 GeV) from a linear accelerator (LINAC) with a heavy metal target (uranium). The principal spallator concept is based on generating fissile fuel for use in LWR nuclear power plants. The spallator functions in conjunction with a reprocessing plant to regenerate and produce the Pu-239 or U-233 for fabrication into fresh LWR reactor fuel elements. Advances in proton accelerator technology has provided a solid base for predicting performance and optimizing the design of a reliable, continuous wave, high-current LINAC required by a fissile fuel production machine.
Date: January 1, 1985
Creator: Steinberg, M.
Partner: UNT Libraries Government Documents Department

Current experimental activities for solid breeder development

Description: The current data base for ceramic breeder materials does not exhibit any negative features as regards to thermophysical, mechanical, and irradiation behavior. All candidate materials show excellent stability for irradiation testing to 3% burnup. In-situ tritium recovery tests show very low tritium inventories for all candidates. Theoretical models are being developed to accurately predict real time release rates. Fabrication of kilogram quantities of materials has been achieved and technology is available for further scale-up.
Date: January 1, 1988
Creator: Johnson, C. E.; Hollenberg, G. W.; Roux, N. & Watanabe, H.
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

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

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

State-of-the-art surveys on sodium-water reaction products cleanup methods and equipment. [LMFBR]

Description: This report describes the basic cleaning methods and equipment which have been used to clean and requalify specimens and components that have been exposed to sodium. Data obtained from laboratory cleaning of test specimens which were earlier exposed to sodium, have been included for various sodium removal methods. Development programs on cleanup methods for removing sodium-water reaction products along with sodium from surfaces and the purification of Intermediate Heat Transport System (IHTS) sodium after emergency dump have been identified.
Date: January 1, 1975
Creator: Anand, R.K.
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

Seismic criteria studies and analyses. Quarterly progress report No. 3. [LMFBR]

Description: Information is presented concerning the extent to which vibratory motions at the subsurface foundation level might differ from motions at the ground surface and the effects of the various subsurface materials on the overall Clinch River Breeder Reactor site response; seismic analyses of LMFBR type reactors to establish analytical procedures for predicting structure stresses and deformations; and aspects of the current technology regarding the representation of energy losses in nuclear power plants as equivalent viscous damping.
Date: January 3, 1975
Partner: UNT Libraries Government Documents Department

U. S. fast reactor materials and structures program

Description: The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program.
Date: January 1, 1984
Creator: Harms, W.O. & Purdy, C.M.
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

Influence of core-assembly refueling requirements on LMFBR core-system design

Description: Liquid metal fast breeder reactor (LMFBR) core assemblies are exposed to an operational environment which induces permanent distortions in their main structural members. These distortions have a substantial impact on core assembly refueling since the distortions are large compared to the available spaces. Core assembly refueling requirements demand that refueling be accomplished without damage to the adjacent core components or to the refueling equipment. This paper describes the core assembly refueling requirements and the design procedures used to demonstrate compliance with the requirements. This paper also provides an assessment relative to the influence of these requirements on LMFBR core system design.
Date: March 1, 1979
Creator: Fox, J.N.
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

Mixing experiments in an alternating wire wrapped assembly

Description: The salt injection experiment was performed in an alternating wire-wrapped, triangular-array bundle to study the coolant mixing behavior pertaining to alternating wrapped wires. Results show that the coolant mixing is much enhanced in the subchannels for this type of bundle compared to the in-phase wire-wrapped bundle currently used in the LMFBR design. A strong interaction between the coolants in the edge and the interior subchannels at four assembly faces (where the alternating wires give 180/sup 0/ out-of-phase configuration) has also been observed. After calibrating the input parameters for the sweeping flow of the COBRA-IIIC/MIT code against this experimental data, the code is employed to predict the coolant temperatures for a skew-powered alternating wire-wrapped blanket assembly. Comparing the code results with the SUPERENERGY code predicted coolant temperature data for an in-phase wrapped assembly under identical operating conditions, it is observed that the hot spot temperature for the alternating wire-wrapped assembly is less than that for the in-phase wire-wrapped bundle by 5 percent (normalized to the hot spot axial temperature rise).
Date: December 1, 1977
Creator: Chiu, C.; Todreas, N.E. & Rohsenow, W.M.
Partner: UNT Libraries Government Documents Department

Status of SACRD: a data base for fast reactor safety computer codes

Description: In 1975 work was initiated to provide a central computerized data collection of evaluated data for use in fast reactor safety computer codes. This data base is called SACRD and is intended to encompass handbook and other nonproblem-dependent data related to LMFBR's, especially at extreme conditions where little or no experimental data are available. Version 1 of the data base was released in the latter part of 1978 and remained the standard version until Version 81, which was released in October 1981.
Date: January 1, 1982
Creator: Greene, N.M.; Flanagan, G.F. & Alter, H.
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