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SM-1A PROJECT--TECHNICAL MANUAL: CHEMISTRY

Description: A manual is given of the equipment and procedures used in the Army Reactor (SM-1A) to control the water purity and makeup. In addition to a description of the primary purification control system, a discussion is presented of the water chemistry control procedures for the auxiliary systems (e.g., the spent-fuel pit, the shield tank, and the waste disposal system). (T.F.H.)
Date: September 29, 1960
Creator: Chupak, J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Small Power Reactor Projects of the United States Atomic Energy Commission

Description: Information on small power reactor projects of the USAEC is summarized. General information concerning the projects as a whole is given. Specific projects discussed include: the Elk River Power Reactor, the Piqua Nuclear Power Facility, the BONUS Power Reactor, the Pathfinder Power Reactor, the small-size pressurized water power reactor, and the experimental low-power process heat reactor. (M.C.G.)
Date: September 28, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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FUEL ELEMENT DEVELOPMENT PROGRAM FOR THE PEBBLE BED REACTOR. Final Report

Description: >The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found-that molecularly deposited'' ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces an… more
Date: April 30, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Dynamic Simulation of Multi-Pass Pressurized Water Nuclear Power Plants by Analog Computer Techniques

Description: A kinetic model of the primary loop of a multi-pass pressurized water reactor power plant is developed to evaluate, by analog computer techniques, the transient response characteristics under conditions of steam generator load and reactor control rod perturbations. Using the 2-pass 28 Mw(t) SM-2 reactor as a typical plant, transient behavior patterns are illustrated and examined for a variety of load inputs, variations in plant constants, and analog model simplifications. (auth)
Date: June 1, 1961
Creator: Brondel, J. O.
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries Government Documents Department
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IN-PILE GAS-COOLED FUEL ELEMENT TEST FACILITY

Description: Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stabilit… more
Date: July 10, 1962
Creator: Zasler, J.; Huntley, W. R.; Gnadt, P. A. & Kress, T. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Survey of Nonmetallic Liquid Coolants for Nuclear-Power Piles

Description: A comprehensive survey of the literature was made in an effort to identify nonmetallic materials of possible usefulness as liquid coolants. Materials having maximum melting points of 1000 deg F and boiling points of l200 deg F were considered, but boiling points above 2200 deg F were preferred. Melting points, boiling points, densities, heat capacities, and thermal conductivities were tabulated. Approximately 190 materials appeared to have melting and boiling temperatures in a suitable range. A… more
Date: May 26, 1950
Creator: Shaw, H.L. & Boulger, F.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Supplementary Study. Extrapolation to Large Central Station Integral Nuclear Superheat Plant

Description: An evaluation was made of the maximum size plant for which the BONUS reactor plant could serve as a realistic prototype and the design changes required to increase the size and characteristics for the present BONUS design such that it could serve as a realistic prototype for the largest feasible integral-superheat reactor power plant. (M.C.G.)
Date: October 31, 1962
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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MODIFICATION OF THE EXPERIMENTAL BOILING WATER REACTOR (EBWR) FOR HIGHER- POWER OPERATION

Description: Supplement to ANL-5607. Alterations and additions made to the Experimental Boiling Water Reactor (EBWR) plant to permit operation at power levels up to I00 Mw(t) are described. Topics covered include over-all system modifications and additions, nuclear component modifications and additions, and reboiler plant component description. (M.C.G.)
Date: April 1, 1962
Creator: Matousek, J.F. comp.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

Description: An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calcula… more
Date: January 1, 1983
Creator: Williams, M.L.; Stallmann, F.W.; Maerker, R.E. & Kam, F.B.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A Mechanism Explaining the Instability of EBR-I, Mark III

Description: Presented at the International Atomic Energy Agencysponsored Seminar on the Physics of Fast and Intermediate Reactors. Vienna, August 3-11, 196l. A feedback model, was developed to account for resonant instabilities in the Mark II core. In this model, the prompt positive power coefficient effect is ascribed to fuel rod bowing and the delayed negative power coefficient effect to thermally lnduced motions in the lower shield plate. Since this model is supported by observations, it is concluded th… more
Date: September 1, 1961
Creator: Smith, R. R.; Matlock, R. G.; McGinnis, F. D.; Novick, M. & Thalgott, F. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Reactor technology. Progress report, January-March 1980

Description: Progress is reported concerning space reactor (SPAR) electric power supply; GCFR reactor safety experiments; structural analysis of HTGR, PWR, and BWR containment vessels and pressure vessels; heat pipe technology development; and nuclear criticality experiments and safety.
Date: June 1, 1980
Creator: Breslow, M. & Sullivan, S. (eds.)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fuel Cycle Program, a Boiling Water Reactor Research and Development Program. Third Quarterly Report, January 1961-March 1961

Description: The continuing analysis of the VBWR core resulted in refinements in the calculations for reactivity in voids, flux leakage and resonance escape probability. The Zircaloy cladding for 25 fuel assemblies was received and passed inspection. Preliminary measurements of VBWR flux oscillations, used to develop instrumentation and data interpretation techniques, showed random normally-distributed oscillations with a predominant frequency of 0.5 to 1.0 cycles/second. A model for analog computer simulat… more
Date: October 31, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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75,000 KILOWATTS OF ELECTRICITY BY NUCLEAR FISSION AT THE HALLAM NUCLEAR POWER FACILITY

Description: For presentation at ASCE Convention in Reno, Nevada on Thursday, June 23, 1860. A description of the Hallam Nuclear Power Facslity is presented. The history of the project, program participants, site description, component development program, reaetor building, reactor structure, reactor core, sodium systems, instrumentation and control, fuel and component handling, auxsilary sustems, special design features, and advantages of sodium graphite reactor systems are discussed. (M.C.G.)
Date: January 1, 1960
Creator: Gronemeyer, F.C. & Merryman, J.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Publications of Los Alamos research 1980

Description: This bibliography is a compilation of unclassified publications of work done at the Los Alamos National Laboratory for 1980. Papers published in 1980 are included regardless of when they were actually written. Publications received too late for inclusion in earlier compilations have also been listed. Declassification of previously classified reports is considered to constitute publication. All classified issuances are omitted-even those papers, themselves unclassified, which were published only… more
Date: September 1, 1981
Creator: Salazar, C. A. & Willis, J. K. (comps.)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2

Description: This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-… more
Date: February 25, 2008
Creator: Winterholler, Kirk
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

Description: The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, … more
Date: September 1, 2008
Creator: Duncan, David S.; Balls, Vondell J. & Austad, Stephanie L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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FABRICATION PROCESS AND PRODUCT QUALITY IMPROVEMENTS IN ADVANCED GAS REACTOR UCO KERNELS

Description: A major element of the Advanced Gas Reactor (AGR) program is developing fuel fabrication processes to produce high quality uranium-containing kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the AGR program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test (“AGR-1) consisted of uranium oxycarbide (UCO) microspheres that werre produced by an internal gelation process followe… more
Date: September 1, 2008
Creator: Barnes, Charles M
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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NGNP Process Heat Utilization: Liquid Metal Phase Change Heat Exchanger

Description: One key long-standing issue that must be overcome to fully realize the successful growth of nuclear power is to determine other benefits of nuclear energy apart from meeting the electricity demands. The Next Generation Nuclear Plant (NGNP) will most likely be producing electricity and heat for the production of hydrogen and/or oil retrieval from oil sands and oil shale to help in our national pursuit of energy independence. For nuclear process heat to be utilized, intermediate heat exchange is … more
Date: September 1, 2008
Creator: Sabharwall, Piyush; Patterson, Mike; Utgikar, Vivek & Gunnerson, Fred
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Hydrogen Production via a Commerically Ready Inorganic membrane Reactor

Description: It has been known that use of the hydrogen selective membrane as a reactor (MR) could potentially improve the efficiency of the water shift reaction (WGS), one of the least efficient unit operations for production of high purity hydrogen from syngas. However, no membrane reactor technology has been reduced to industrial practice thus far, in particular for a large-scale operation. This implementation and commercialization barrier is attributed to the lack of a commercially viable hydrogen selec… more
Date: June 30, 2007
Creator: Liu, Paul
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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RESULTS OF TESTS TO DEMONSTRATE A SIX-INCH-DIAMETER COATER FOR PRODUCTION OF TRISO-COATED PARTICLES FOR ADVANCED GAS REACTOR EXPERIMENTS

Description: The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory’s (INL’s) Advanced Test Reactor. TRISOcoated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a two inch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial s… more
Date: September 1, 2008
Creator: Barnes, Charles M
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

Description: The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept’s inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nod… more
Date: September 1, 2008
Creator: Hiruta, Hikaru; Ougouag, Abderrafi M.; Gougar, Hans D. & Ortensi, Javier
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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