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HTGR Generic Technology Program: materials technology reactor; operating experience; medium-enriched-uranium fuel development. Quarterly progress report for the period ending July 31, 1978

Description: The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.
Date: August 1, 1978
Partner: UNT Libraries Government Documents Department

Brayton Isotope Power System (BIPS) superalloy ground demonstration system (GDS-S) detail design review

Description: The material presented at the GDS-S design review meeting held at Phoenix, Arizona on September 7 and 8, 1977, is reported. This design review was specifically scoped to examine the Hastelloy-X Heat Source Heat Exchanger (HSHX) and the Hastelloy-X bellows. These are new components required to upgrade the Workhorse Loop (WHL) to the GDS configuration. Additional topics covered in support of those items were: reliability; materials; and the WHL-to-GDS conversion sequence.
Date: September 12, 1977
Partner: UNT Libraries Government Documents Department

Study of tertiary creep instability in several elevated-temperature structural materials

Description: Data for a number of common elevated temperature structural materials have been analyzed to yield mathematical predictions for the time and strain to tertiary creep at various rupture lives and temperatures. Materials examined include types 304 and 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, alloy 718, Hastelloy alloy X, and ERNiCr--3 weld metal. Data were typically examined over a range of creep temperatures for rupture lives ranging from less than 100 to greater than 10,000 hours. Within a given material, trends in these quantities can be consistently described, but it is difficult to directly relate the onset of tertiary creep to failure-inducing instabilities. A series of discontinued tests for alloy 718 at 649 and 620/sup 0/C showed that the material fails by intergranular cracking but that no significant intergranular cracking occurs until well after the onset of tertiary creep.
Date: January 1, 1978
Creator: Booker, M.K. & Sikka, V.K.
Partner: UNT Libraries Government Documents Department

Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys. [HTGR]

Description: Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650/sup 0/C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTGR materials to more prototypic, low concentrations of I/sub 2/, Te/sub 2/ and CsI in the presence of low partial pressures of O/sub 2/. These samples are scheduled for mechanical bend tests after exposure to determine the degree of embrittlement.
Date: January 1, 1978
Creator: Aronson, S.; Chow, J.G.Y.; Soo, P. & Friedlander, M.
Partner: UNT Libraries Government Documents Department

RADIATION EFFECTS ON FAST REACTOR CLADDING AND STRUCTURAL MATERIALS.

Description: The effect of radiation on the time-, temperature-, and stress- dependent properties of selected heat-resistant alloys and refractory metals was determined. Causes of property changes were identified and remedial measures are planned. Materials investigated include A-286 Iron alloy; Hastelloy X; Hastelloy N; Hastelloy R-235; Al - Cr - Fe - Y alloys; AlSl 304, 316, and 348 stainless stells, ASTM-A 302B and A 350-LF3 pressure vessel steels, and various Inconel and Incoloy alloys and heat-resisting metals such as Mo, Nb, Ta, V, and W. (F.S.)
Date: October 31, 1969
Creator: Moteff, J.; Kingsbury, F.D. & Smith, J.P.
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

Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

Description: The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.
Date: March 7, 1980
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor helium compatibility studies: results of 10,000-hour exposure of selected alloys in simulated reactor helium

Description: Work on the HTGR Helium Compatibility Task accomplished during the period March 31, 1977 through September 30, 1979, is documented in this report. Emphasis is on the results and analyses of creep data to 10,000 h and the detailed metallurgical evaluations performed on candidate alloy specimens tested for up to 10,000 h. Long-term creep and unstressed aging data in controlled-impurity helium and in air at 800, 900, and 1000/sup 0/C are reported for alloys included in the program in FY-76, including the wrought solid-solution-strengthened alloys, Hastelloy X, Hastelloy S, RA 333, and HD 556, and the centrifugally cast austenitic alloys, HK 40, Supertherm, Manaurite 36X, Manaurite 36XS, and Manaurite 900.
Date: May 1, 1980
Creator: Lechtenberg, T.A.; Stevenson, R.D. & Johnson, W.R.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

Description: During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.
Date: June 1, 1983
Creator: Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

Description: Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.
Date: November 24, 1978
Partner: UNT Libraries Government Documents Department