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Examination of compression and resilience characteristics of fibrous insulation blankets

Description: Load-deflection characteristics of alumina and alumino-silicate fibrous blankets were experimentally determined. Load retention and springback capability of combinations of these materials were measured in a 10,000-hour test at surface temperatures of 650 to 1000/sup 0/C (1200 to 1832/sup 0/F). Experimental results are presented and future testing plans are discussed.
Date: August 1, 1979
Creator: Brislin, R.J. & Middleton, A.
Partner: UNT Libraries Government Documents Department

Application of programmable controllers to vacuum system interlocks

Description: This paper describes the Doublet III Vacuum Control System in which all input signals and output loads are connected to a programmable controller (PC) for logical interfacing. Input signals derived from CAMAC, control panels, limit switches, etc., are implemented as output signals to CAMAC, vacuum valves, pump motors, etc., according to a logic program stored in the PC memory. The memory can be easily programmed by anyone familar with either Boolean algebra or relay-ladder network diagrams. The program data is entered with the aid of a calculator like, keyboard instrument with LED readout displays. The PC system contains a 1024 word RAM memory with a battery backup system to provide 72 hours protection of contents in case of power failure. (MOW)
Date: November 1, 1979
Creator: Lee, G. & Moore, D.
Partner: UNT Libraries Government Documents Department

Effects of surface condition on the corrosion of candidate structural materials in a simulated HTGR-GT environment

Description: A simulated high-temperature gas-cooled reactor (HTGR) helium environment was used to study the effects of surface finish conditions on the subsequent elevated-temperature corrosion behavior of key candidate structural materials. The environment contained helium with 500 ..mu..atm H/sub 2//50 ..mu..atm CO/50 ..mu..atm CH/sub 4//<0.5 ..mu..atm H/sub 2/O at 900/sup 0/C with total test exposure durations of 3000 hours. Specimens with lapped, grit-blasted, pickled, and preoxidized surface conditions were studied. Materials tested included two cast superalloys, IN 100 and IN 713LC; one centrifugally cast high-temperature alloy, HK 40 one oxice-dispersion-strengthened alloy, Inconel MA 754; and three wrought high-temperature alloys, Hastelloy Alloy X, Inconel Alloy 617, and Alloy 800H.
Date: February 1, 1980
Creator: Thompson, L.D.
Partner: UNT Libraries Government Documents Department

Effects of the HTGR-gas turbine on national reactor strategies

Description: A specific role for the HTGR in a national energy strategy is examined. The issue is addressed in two ways. First, the role of the HTGR-GT Binary cycle plant is examined in a national energy strategy based on symbiosis between fast breeder and advanced converter reactors utilizing the thorium U233 fuel cycle. Second, the advantages of the HTGR-GT dry-cooled plant operating in arid regions is examined and compared with a dry-cooled LWR. An event tree analysis of potential benefits is applied.
Date: November 1, 1979
Creator: Ligon, D.M. & Brogli, R.H.
Partner: UNT Libraries Government Documents Department

Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

Description: Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment.
Date: October 1, 1979
Creator: Kang, C.S. & Torri, A.
Partner: UNT Libraries Government Documents Department

Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

Description: This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.
Date: May 1, 1980
Creator: Barsell, A.W.
Partner: UNT Libraries Government Documents Department

Pebble Bed Reactor review update. Fiscal year 1979 annual report

Description: Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability.
Date: January 1, 1980
Partner: UNT Libraries Government Documents Department

Line focus solar central power systems. Phase I. Final report, September 30, 1978-October 31, 1979

Description: A conceptual design study was performed of a stand-alone Line Focus Solar Central Power System based on the fixed mirror solar concentrator (FMSC) for heat collection and draw salt (a 50% molar mixture of sodium nitrate and potassium nitrate) for heat transport and storage. Parametric analyses were performed at the subsystem level, and models were developed that were employed in a computerized simulation to minimize the cost of electricity (COE) by adjusting system design parameters. A design was prepared and costed for a first commercial plant with a rating of 100 MW(e) and a storage capacity equivalent to 420 MW(e)-hr of generation. The resulting plant achieves an annual capacity of 45.6%. Scaling studies indicate reductions in the COE for increased capacity factor and increased plant rating. Assessments of the plant concept indicate it should be acceptable to utilities on the basis of technical and operational considerations, but that reductions from the first 100-MW(e) plant cost would be required to achieve substantial market penetration.
Date: September 1, 1979
Partner: UNT Libraries Government Documents Department

Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

Description: A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

HTGR gas turbine program. Semiannual progress report, April 1-September 30, 1978

Description: This report describes work performed under the gas turbine HTGR (HTGR-GT) program, Department of Energy Contract DE-AT03-76-SF70046, during the period April 1, 1978 through September 30, 1978. The work reported covers the demonstration and commercial plant concept studies including plant layout, heat exchanger studies, turbomachine studies, systems analysis, and reactor core engineering.
Date: December 1, 1979
Partner: UNT Libraries Government Documents Department

HTGR Gas Turbine Program. Semiannual progress report for the period ending September 30, 1979

Description: Information on the HTGR-GT program is presented concerning systems design methods; systems dynamics methods; alternate design; miscellaneous controls and auxiliary systems; structural mechanics; shielding analysis; licensing; safety; availability; reactor turbine system integration with plant; PCRV liners, penetrations, and closures; PCRV structures; thermal barrier; reactor internals; turbomachinery; turbomachine remote maintenance; control valve; heat exchangers; plant protection system; and plant control system.
Date: May 1, 1980
Partner: UNT Libraries Government Documents Department

Hybrid reactor safety study. Annual report, October 1, 1978-September 30, 1979

Description: A preliminary generic safety evaluation of the fusion-fission hybrid reactor concept has been performed and a hybrid reactor safety program plan for guiding future safety work has been proposed. The emphasis of the work was limited to accident analysis where the main concern is for the health and safety of the public. Major radioactive sources in the hybrid were identified and their inventories compared to those of fission reactors. The means for accidental release of radioactivity to the public were identified, as were the barriers which preclude such accidental releases. Consequence analyses of hypothetical bounding accidents potentially defining the upper bound envelope of risk/consequence to the population and environment surrounding the hybrid site were performed.
Date: December 1, 1979
Partner: UNT Libraries Government Documents Department

HTGR gas-turbine program. Semiannual progress report for period ending March 31, 1980

Description: This report describes the conceptual design and analysis performed by General Atomic Company and its subcontractors for the US Department of Energy on the direct cycle gas turbine high-temperature gas-cooled reactor. The primary accomplishments for this period were cost reduction studies, turbomachinery failure analysis, and alternate plant concept evaluation.
Date: January 1, 1982
Partner: UNT Libraries Government Documents Department

HTGR process heat program design and analysis. Final report, FY-79

Description: This report summarizes the results of concept design studies at General Atomic Company during FY-79 for an 842-MW(t) Very High Temperature Reactor (VHTR) utilizing an intermediate helium heat transfer loop to provide thermal energy for the production of hydrogen or reducing gas (H/sub 2/ + CO) by steam-reforming of a light hydrocarbon. Basic carbon sources may be coal, residual oil, or oil shale. The report summarizes conceptual design tasks conducted on the prestressed concrete reactor vessel, thermal barrier, intermediate heat exchanger, reformer, and steam generator. The substantial completion of first generation programming for a performance/optimization code and the preparation of a topical safety report and other safety evaluation studies are reported. The completion of balance of plant criteria specifications and a balance of plant cost estimate is also reported.
Date: December 1, 1979
Partner: UNT Libraries Government Documents Department

Process heat reactor design and analaysis. Quarterly progress report, April 1-June 30, 1978

Description: This report summarizes the third quarter FY-1978 results of concept design studies at General Atomic Company (GA) for an 842-MW(t) VHTR utilizing an intermediate helium heat transfer loop to provide thermal energy for the production of reducing gas (H/sub 2/ + CO) by steam-reforming a light hydrocarbon. Basic carbon sources may be coal, residual, or oil shale. The report summarizes the various plant configurations selected for the study and presents the conceptual plant layout drawings. Results of design studies on the intermediate heat exchanger are also presented. The status of the performance/optimization code development is discussed, and completion of the core auxiliary cooling system study is summarized.
Date: June 30, 1978
Partner: UNT Libraries Government Documents Department

Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

Description: The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.
Date: May 1, 1980
Partner: UNT Libraries Government Documents Department

3000 MW(t) HTGR - gas turbine non-intercoolel. Technical evaluation report

Description: This report summarizes all the technical work performed on the 3000-MW(t) 3-loop High-Temperature Gas-Cooled Reactor Gas Turbine design as of June 1979. Although the plant configuration has changed to a 2000-MW(t) 2-loop plant, most of the technical assessments described in this report are still applicable to the 2000-MW(t) plant. The report covers the criteria under which the plant was designed, the technical feasibility problems associated with the plant and their potential solutions, and other potential applications and improvements which could make the gas turbine concept more attractive economically.
Date: December 1, 1979
Partner: UNT Libraries Government Documents Department

HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1979

Description: The technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-79 are reported. The report covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop an MEU fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant.
Date: November 1, 1979
Partner: UNT Libraries Government Documents Department

Development of a surveillance robot for dimensional and visual inspection of fuel and reflector elements from the Fort St. Vrain HTGR

Description: A robotic device has been developed for dimensional and visual inspection of irradiated HTGR core components. The robot consists of a rotary table and a two-finger probe, driven by stepping motors, and four remotely controlled television cameras. Automated operation is accomplished via minicomputer control. A total of 51 irradiated fuel and reflector elements were inspected at a fraction of the time and cost required for conventional methods.
Date: November 1, 1979
Creator: Wallroth, C.F.; Marsh, N.I.; Miller, C.M.; Saurwein, J.J. & Smith, T.L.
Partner: UNT Libraries Government Documents Department

Characterization of thermally sprayed coatings for high-temperature wear-protection applications

Description: Under normal high-temperature gas-cooled reactor (HTGR) operating conditions, faying surfaces of metallic components under high contact pressure are prone to friction, wear, and self-welding damage. Component design calls for coatings for the protection of the mating surfaces. Anticipated operating temperatures up to 850 to 950/sup 0/C (1562 to 1742/sup 0/F) and a 40-y design life require coatings with excellent thermal stability and adequate wear and spallation resistance, and they must be compatible with the HTGR coolant helium environment. Plasma and detonation-gun (D-gun) deposited chromium carbide-base and stabilized zirconia coatings are under consideration for wear protection of reactor components such as the thermal barrier, heat exchangers, control rods, and turbomachinery. Programs are under way to address the structural integrity, helium compatibility, and tribological behavior of relevant sprayed coatings. In this paper, the need for protection of critical metallic components and the criteria for selection of coatings are discussed. The technical background to coating development and the experience with the steam cycle HTGR (HTGR-SC) are commented upon. Coating characterization techniques employed at General Atomic Company (GA) are presented, and the progress of the experimental programs is briefly reviewed. In characterizing the coatings for HTGR applications, it is concluded that a systems approach to establish correlation between coating process parameters and coating microstructural and tribological properties for design consideration is required.
Date: March 1, 1980
Creator: Li, C.C.
Partner: UNT Libraries Government Documents Department

Core seismic methods verification report. [HTGR]

Description: Information on HTGR reactor core seismic requirements is presented concerning element properties and code parameters; correlation and verification of the codes; sensitivity studies; and application to design.
Date: December 1, 1979
Creator: Olsen, B.E.; Shatoff, H.D.; Rakowski, J.E.; Rickard, N.D.; Thompson, R.W.; Tow, D. et al.
Partner: UNT Libraries Government Documents Department

Core seismic methods verification report

Description: This report presents the description and validation of the analytical methods for calculation of the seismic loads on an HTGR core and the core support structures. Analytical modeling, integration schemes, parameter assignment, parameter sensitivity, and correlation with test data are key topics which have been covered in detail. Much of the text concerns the description and the results of a series of scale model tests performed to obtain data for code correlation. A discussion of scaling laws, model properties, seismic excitation, instrumentation, and data reduction methods is also presented, including a section on the identification and calculation of statistical errors in the test data.
Date: December 1, 1979
Creator: Olsen, B.E.; Shatoff, H.D.; Rakowski, J.E.; Rickard, N.D.; Thompson, R.W.; Tow, D. et al.
Partner: UNT Libraries Government Documents Department