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Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

Description: The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department
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Overview of US fast-neutron facilities and testing capabilities

Description: Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected.
Date: January 1, 1982
Creator: Evans, E.A.; Cox, C.M. & Jackson, R.J.
Partner: UNT Libraries Government Documents Department
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Design and operation of a rapid thermal transient component testing sodium loop

Description: A specific problem developed during the design of an on-line sampling system for the Sodium Loop Safety Facility fast breeder reactor experiments. Rapid fluctuations in the sodium temperature, caused by reactor operation and shutdown, exposed the system components to conditions that could result in fatigue failure. A component test loop was designed and built at the Idaho National Engineering Laboratory to allow experimental qualification of component integrity. This paper outlines test system requirements, describes the design and special features, presents test procedures ad relates significant operating experience.
Date: April 16, 1984
Creator: Crandall, D. L.
Partner: UNT Libraries Government Documents Department
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United States Department of Energy breeder reactor staff training domestic program

Description: Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described.
Date: January 1, 1984
Partner: UNT Libraries Government Documents Department
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Conclusions Drawn From Subcritical Multiplication Results in ZPPR

Description: Modified source multiplication (MSM) has been investigated as a measurement technique for several LMFBR designs in the 1000 MWt class. This investigation has been based on the results of hundreds of measurements in ZPPR as well as on detailed analysis of some of the components of the method. The results of this study have special significance for critical-experiment planning and for the use of ex-vessel detectors in operating LMFBRs. The calculated factors that are applied in the MSM technique for ex-core detectors were found to be particularly insensitive to the calculational method. This apparently resulted from two facts: (1) the calculated factors are ratios of ratios, which reduce the sensitivity to first order corrections, and (2) most neutrons which cause reactions in an ex-core detector originate in the small volume of core nearest to the detector. Maps of key parameters in the MSM technique have been generated to promote a general understanding of the method and to assure its proper application.
Date: January 1, 1978
Creator: Carpenter, S. G.; McFarlane, H. F.; Lineberry, M. J. & Beck, C. L.
Partner: UNT Libraries Government Documents Department
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Analysis of the TREAT loss-of-flow tests L6 and L7 using SAS3D

Description: The TREAT loss-of-flow tests L6 and L7 have been analyzed using the SAS3D accident analysis code. The impetus for the analysis was the need for experimentally supported fuel motion modeling in whole core accident studies performed in support of licensing of the Clinch River Breeder Reactor Project. The input prescription chosen for the SAS3D/SLUMPY fuel motion model gave reasonable agreement with the test results. Tests L6 and L7, each conducted with a cluster of three fuel pins, were planned to simulate key events in the loss-of-flow accident scenario for the Clinch River homogeneous reactor.
Date: January 1, 1985
Creator: Morris, E. E.; Simms, R. & Gruber, E. E.
Partner: UNT Libraries Government Documents Department
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Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

Description: A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.
Date: October 1, 1985
Creator: Adams, S. R.
Partner: UNT Libraries Government Documents Department
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Instrument response during overpower transients at TREAT

Description: A program to empirically analyze data residuals or noise to determine instrument response that occurs during in-pile transient tests is out-lined. As an example, thermocouple response in the Mark III loop during a severe overpower transient in TREAT is studied both in frequency space and in real-time. Time intervals studied included both constant power and burst portions of the power transient. Thermocouple time constants were computed. Benefits and limitations of the method are discussed.
Date: January 1, 1982
Creator: Meek, C. C.; Bauer, T. H.; Hill, D. J.; Froehle, P. H.; Klickman, A. E.; Tylka, J. P. et al.
Partner: UNT Libraries Government Documents Department
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Fuel-cycle costs for alternative fuels

Description: This paper compares the fuel cycle cost and fresh fuel requirements for a range of nuclear reactor systems including the present day LWR without fuel recycle, an LWR modified to obtain a higher fuel burnup, an LWR using recycle uranium and plutonium fuel, an LWR using a proliferation resistant /sup 233/U-Th cycle, a heavy water reactor, a couple of HTGRs, a GCFR, and several LMFBRs. These reactor systems were selected from a set of 26 developed for the NASAP study and represent a wide range of fuel cycle requirements.
Date: January 1, 1980
Creator: Rainey, R.H.; Burch, W.D.; Haire, M.J. & Unger, W.E.
Partner: UNT Libraries Government Documents Department
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Pump, sodium, inducer, intermediate size (ISIP) (impeller/inducer/diffuser retrofit)

Description: This specification defines the requirements for the Intermediate-Size Inducer Pump (ISIP), which is to be made by replacing the impeller of the FFTF Prototype Pump with a new inducer, impeller, diffuser, seal, and necessary adapter hardware. Subsequent testing requirements of the complete pump assembly are included.
Date: April 21, 1978
Creator: Paradise, D.R.
Partner: UNT Libraries Government Documents Department
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Status of gamma-ray heating characterization in LMFBR

Description: Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR.
Date: November 1, 1975
Creator: Gold, R.
Partner: UNT Libraries Government Documents Department
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Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

Description: The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.
Date: January 1, 1985
Creator: Morris, E. E.; Dunn, F. E.; Simms, R. & Gruber, E. E.
Partner: UNT Libraries Government Documents Department
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Automated start-up of EBR-II: A preview

Description: Oak Ridge National Laboratory (ORNL) and Argonne National Laboratory (ANL) are undertaking a joint project to develop control philosophies, strategies, and algorithms for computer control of the start-up mode of the Experimental Breeder Reactor II (EBR-II). The major objective of this project is to show that advanced liquid-metal reactor (LMR) plants can be operated from low power to full power using computer control. Development of an automated control system with this objective in view will help resolve specific issues and provide proof through demonstration that automatic control for plant start-up is feasible. This paper describes the approach that will be used to develop such a system and some of the features it is expected to have. Structured, rule-based methods, which will provide start-up capability from a variety of initial plant conditions and degrees of equipment operability, will be used for accomplishing mode changes during plant start-up. Several innovative features will be incorporated such as signal, command, and strategy validation to maximize reliability, flexibility to accommodate a wide range of plant conditions, and overall utility. Continuous control design will utilize figures of merit to evaluate how well the controller meets the mission requirements. The operator interface will have unique ''look ahead'' features to let the operator see what will happen next. 15 refs., 7 figs., 1 tab.
Date: January 1, 1989
Creator: Kisner, R. A.
Partner: UNT Libraries Government Documents Department
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Large EM pump program review. [Annular linear induction]

Description: Objectives were to review the Annular Linear Induction Pump (ALIP) concept and program status, determine the desirability for such a concept for large scale plant Main Heat Transport systems, and establish the extent of interest and support for preceeding with development.
Date: January 23, 1976
Partner: UNT Libraries Government Documents Department
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Liquid-metal pumps for large-scale breeder-reactor plant (prototype pump)

Description: This report presents the recommended pump design for use in Large Scale Liquid Metal Fast Breeder Reactor plants. The base design for the pump will circulate 127,000 GPM of liquid sodium at temperatures up to 850/sup 0/F and with a total discharge head at the design point of 500 feet Na with an impeller that is 40 feet below the sodium seal. The pump design is predicated on developing an impeller design which will have a suction specific speed (S/sub n/) of about 20,000 with 20 feet NPSH available, which will result in a pump speed of 530 RPM at design conditions. The design is based on the technology developed in the design and fabrication of FFTF pumps, the design efforts for the Clinch River Breeder Reactor Pump design study and other technology.
Date: July 1, 1976
Creator: Lindsay, M. (comp.)
Partner: UNT Libraries Government Documents Department
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Improved algorithms for the calculation of resolved resonance cross sections with applications to the structural Doppler effect in fast reactors

Description: Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC/sup 2/-2.
Date: October 1, 1980
Creator: Hwang, R. N.; Toppel, B. J. & Henryson II, H.
Partner: UNT Libraries Government Documents Department
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Fast reactor operation in the United States

Description: Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO/sub 2/-PuO/sub 2/ ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed.
Date: January 1, 1978
Creator: Smith, R. R. & Cissel, D. W.
Partner: UNT Libraries Government Documents Department
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SLSF local fault safety experiment P4: summary and conclusions. [Sodium Loop Safety Facility]

Description: Sodium Loop Safety Facility (SLSF) experiment P4 in ETR was performed to investigate the consequences of an upper-bound or worse-than-worst case local fault configuration. P4 was intended to bound the consequences of credible subassembly faults by ejecting molten fuel into a 37-pin bundle of full-length Fast Test Reactor (FTR)-type pins and failing fuel with the potential for further cladding and fuel-pin damage. In addition to ejecting a large amount of molten fuel at or near full power, experiment objectives were to evaluate the severity of molten fuel-coolant interactions (MFCIs) and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors in time to prevent significant fuel failure propagation.
Date: January 1, 1985
Creator: Thompson, D. H.; Ragland, W. A.; Holland, J. W.; Dever, D. J.; Braid, T. H.; Baldwin, R. D. et al.
Partner: UNT Libraries Government Documents Department
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Coded aperture imaging of reactor illuminated LMFBR fuel pins

Description: As a part of Sandia Laboratories program to study simulated core disruptive accidents in reactor safety research, a fuel motion detection system based on coded aperture imaging is being developed. Until the experiments reported here were conducted it was not known whether either of two important questions could be properly answered. First, could an actual fuel pin illuminated by a reactor be imaged with a coded aperture and an active detector system with acceptable frame rates and spatial resolution. Second, could collimator and shield structures be fabricated for the Annular Core Pulsed Reactor (ACPR), for which the system is being built, that would provide acceptable signal to background ratios. Both of these questions can now be answered in the affirmative because of the favorable experiments recently conducted at the Sandia Pulsed Reactor (SPR II) and at the ACPR.
Date: January 1, 1977
Creator: Kelly, J. G. & Stalker, K. T.
Partner: UNT Libraries Government Documents Department
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Hydrodynamic analysis of the LMFBR prompt burst excursion (PBE) experiment

Description: A series of in-pile experiments has been conducted at Sandia Laboratories to provide information on pressure levels and conversion of thermal energy into mechanical work in LMFBR cores during hypothetical, superprompt-critical excursions. Pressures generated in these experiments are recorded by a pressure transducer located at the top and bottom of a sodium channel surrounding a single, fresh UO/sub 2/ fuel pin. Work energy conversion is measured by a linear motion transducer connected to a piston at the top of the sodium column. Since the pressure transducers are located fairly far from the location of pin failure, it becomes necessary to determine the effect of channel geometry and piston motion on the observed pressure data. A two-dimensional, hydrodynamic analysis of pressure pulse propagation in the fuel pin-coolant channel geometry was therefore performed using the CSQII computer code. The initial series of PBE experiments consists of single, fresh UO/sub 2/ pins surrounded by a sodium-filled or dry-coolant channel contained in a closed test capsule. The capsule is subjected to a maximum pulse in the Annular Core Pulse Reactor (ACPR) resulting in an energy deposition of from 2350 to 2900 J/g (14 and 20 percent enriched pins). The pulse width at half maximum (PWHM) is about 5 ms.
Date: January 1, 1977
Creator: Young, M. F.
Partner: UNT Libraries Government Documents Department
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Prompt burst energetics experiments: fresh oxide/sodium series

Description: A series of in-pile experiments has been performed to provide information on thermal energy to work conversion under prompt burst excursion (PBE) conditions. These consisted of single pin tests using fresh uranium oxide or uranium carbide fuel in a capsule geometry, with either stagnant sodium or helium in the coolant channel. The experiments were irradiated with single or double pulses in the Annular Core Pulse Reactor (ACPR) to provide energy depositions up to 2900 J/g. This report covers the seven single and five double pulse UO/sub 2/ sodium-in tests. Experimental data includes pressure and linear motion transducer histories, measured work-energy conversion efficiencies, and post-irradiation examination. Analysis includes derived work-energy conversion efficiencies (up to 0.54%), pin failure modeling, hydrodynamic analysis of pressure pulse propagation in the channel, and piston stopping effects. Initial pressure events in the single pulse experiments appear to be dominated by fuel vapor pressure. Definite fuel-coolant interactions were observed in several experiments, including some that were coincident with stopping of the linear motion transducer piston, suggesting a possible triggering effect by the deceleration pressure.
Date: August 1, 1978
Creator: Reil, K.O. & Young, M.F.
Partner: UNT Libraries Government Documents Department
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Experimental Breeder Reactor-II dynamic modeling and code verification

Description: The Experimental Breeder Reactor No. 2 (EBR-II) has been modeled using a recently developed special purpose block oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP). The purpose of the work was to develop and verify the code and use it to support the Operational Reliability Test Program at EBR-II. Designed to be set up directly from block diagrams of the reactor system, DSNP allows easy interchange of modules which simulate individual components of the plant with differing levels of complexity.
Date: June 1, 1983
Creator: Lehto, W. K.; Dean, E. M.; Larson, H. A. & Koenig, J. F.
Partner: UNT Libraries Government Documents Department
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