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Thermal stability of high temperature structural alloys

Description: High temperature structural alloys were evaluated for suitability for long term operation at elevated temperatures. The effect of elevated temperature exposure on the microstructure and mechanical properties of a number of alloys was characterized. Fe-based alloys (330 stainless steel, 800H, and mechanically alloyed MA 956), and Ni-based alloys (Hastelloy X, Haynes 230, Alloy 718, and mechanically alloyed MA 758) were evaluated for room temperature tensile and impact toughness properties after exposure at 750 C for 10,000 hours. Of the Fe-based alloys evaluated, 330 stainless steel and 800H showed secondary carbide (M{sub 23}C{sub 6}) precipitation and a corresponding reduction in ductility and toughness as compared to the as-received condition. Within the group of Ni-based alloys tested, Alloy 718 showed the most dramatic structure change as it formed delta phase during 10,000 hours of exposure at 750 C with significant reductions in strength, ductility, and toughness. Haynes 230 and Hastelloy X showed significant M{sub 23}C{sub 6} carbide precipitation and a resulting reduction in ductility and toughness. Haynes 230 was also evaluated after 10,000 hours of exposure at 850, 950, and 1050 C. For the 750--950 C exposures the M{sub 23}C{sub 6} carbides in Haynes 230 coarsened. This resulted in large reductions in impact strength and ductility for the 750, 850 and 950 C specimens. The 1050 C exposure specimens showed the resolution of M{sub 23}C{sub 6} secondary carbides, and mechanical properties similar to the as-received solution annealed condition.
Date: March 1, 1999
Creator: Jordan, C.E.; Rasefske, R.K. & Castagna, A.
Partner: UNT Libraries Government Documents Department

Testing of candidate materials for their resistance to alkali-vapor adsorption in PFBC and gasification environments. Final report

Description: Laboratory-scale studies were performed to identify metallic material(s) having no, or limited, affinity for alkali vapors in an environment of either the off-gas from pressurized fluidized-bed combustion (PFBC) or the fuel gas from coal gasification. Such materials would be potential candidates for use as components in advanced coal-utilization systems. The following materials were tested for adsorption of NaCl vapor at 870--875 C and atmospheric pressure in a simulated PFBC off-gas (oxidizing) doped with 80 ppmW NaCl vapor: iron-based Type 304 stainless steel (304 SS), nickel-based Hastelloy C-276 and Hastelloy X alloys, cobalt-based Haynes No. 188 alloy, noble-metal-coated 304 SS, aluminized 304 SS, and ZrO{sub 2}-coated 304 SS. The Haynes No. 188 alloy and the aluminized 304 SS were also tested for their NaCl-vapor adsorption in a simulated gasification fuel gas (reducing) under the same test conditions as in the PFBC off-gas test. After 100 h of testing, the specimens were analyzed with a SEM equipped with an energy dispersive X-ray analyzer, and by an AES. The aluminized 304 SS had the least tendency to adsorb NaCl vapor, as well as an excellent resistance to corrosion as a result of the formation of a protective layer of Al{sub 2}O{sub 3} on its surface. In the reducing environment, however, the aluminized 304 SS was badly corroded by H{sub 2}S attack. The Haynes No. 188 showed virtually no NaCl-vapor adsorption and only limited H{sub 2}S attack. The authors recommend further long-term parametric studies to quantitate alkali-vapor adsorption as a function of operating variables for (1) the aluminized 304 SS in the PFBC off-gas environment and (2) the Haynes No. 188 in the gasification fuel gas environment.
Date: August 1, 1995
Creator: Lee, S.H.D.; Natesan, K. & Swift, W.M.
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

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development program. Progress report, October 1, 1981-December 31, 1981. [Alloy-MA-956; alloy-MA-754]

Description: Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is descibed; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/, and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in the controlled-purity helium in the intensive screening test program is discussed. The results of metallographic studies of screening alloys exposed in controlled-purity helium for 3000 hours at 750/sup 0/C and 5500 hours at 950/sup 0/C, 3000 hours at 1050/sup 0/C and 6000 hours at 1050/sup 0/C and for weldments exposed in controlled-purity helium for 6000 hours at 750/sup 0/C and 6000 hours at 1050/sup 0/C are presented and discussed.
Date: June 15, 1982
Creator: Kimball, O.F.
Partner: UNT Libraries Government Documents Department

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

Description: Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. 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. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.
Date: June 25, 1980
Partner: UNT Libraries Government Documents Department

Cobalt-60 heat source demonstration program. Phase III. Fabrication. Final report

Description: Significant accomplishments completed during Phase III of the /sup 60/Co Heat Source Demonstration program include the following: encapsulation of 2 MCi of /sup 60/Co; fabrication of the heat source, including the ASME coded pressure vessel/core assembly, and biological shielding; endurance testing of a prototype heat pipe for a period of 28 months; fabrication and qualification of the heat pipe emergency cooling subsystem; issue of the safety evaluation report, reference 3, and the operations manual, reference 4; and heat source assembly. The planned demonstration test program was modified to include testing of a total power system. Based on an evaluation of available power conversion systems, which included the closed-cycle Brayton and organic Rankine systems, the closed-cycle Brayton system was selected for use. Selection was based on advantages offered by the direct coupling of this conversion system with the gas-cooled heat source. In implementing the test program, the AiResearch BCD power conversion system was to be coupled to the heat source following initial heat source performance testing and part way through the endurance test. In accordance with the program redirection the following Phase IV checkout operations were completed to evaluate procedural and hardware acceptability: heat source dummy fueling; fueling cask sielding survey; and heat source shielding survey (single pin). Completion of these latter activities verified the acceptability of critical characteristics of the heat source and its supporting equipment.
Date: June 1, 1973
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

ROVER ROCKET MATERIALS.

Description: No Description Available.
Date: October 31, 1969
Creator: Clausing, R.E.
Partner: UNT Libraries Government Documents Department

CLADDING MATERIALS FOR SNAP-8.

Description: No Description Available.
Date: October 31, 1969
Creator: Weir, J.R. Jr.; McCoy, H.E. Jr. & Harman, D.G.
Partner: UNT Libraries Government Documents Department

Testing of bellows for engineering systems. Part 1. [For Brayton Isotope Power System]

Description: Bellows made of Hastelloy X and Type 347 stainless steel will be used in the Brayton Isotope Power System (BIPS). Development of this system will involve the Work Horse Loop (WHL) and the Ground Demonstration System (GDS). The primary function of these bellows is to accommodate relative movement of parts of the system, which results from differential temperatures. Methods for performing several types of tests on metal bellows have been developed. Three Hastelloy X bellows were subjected to low-cycle fatigue tests, and their failure times were very close to those predicted by the designer. The failures were largely intergranular and typical of the type expected at elevated temperatures.
Date: July 1, 1979
Creator: McCoy, H.E. Jr.
Partner: UNT Libraries Government Documents Department

Gas-cooled reactor programs: High-Temperature Gas-cooled Reactor Base-Technology Program. Annual progress report for period ending December 31, 1978

Description: Progress in HTGR studies is reported in the following areas: fission product transport and coolant impurity effects, fueled graphite development, PCRV development, structural materials, characterization and standardization of graphite, and evaluation of the pebble-bed type HTGR.
Date: June 1, 1979
Creator: Homan, F.J. & Kasten, P.R.
Partner: UNT Libraries Government Documents Department

Creep and tensile properties of alloy 800H-Hastelloy X weldments. [HTGR]

Description: Hastelloy X and alloy 800H were joined satisfactorily by the gas tungsten arc welding process with ERNiCr-3 filler and the shielded metal arc welding process with Inco Weld A filler. Test specimens were of two types: (1) made entirely of deposited Inco Weld A and (2) machined transverse across the weldments to include Hastelloy X, filler metal (ERNiCr-3 or Inco Weld A), and alloy 800H. They were aged 2000 and 10,000 h and subjected to short-term tensile and creep tests. Inco Weld A and ERNiCr-3 are both suitable filler metals and result in welds that are stronger than the alloy 800H base metal.
Date: August 1, 1983
Creator: McCoy, H. E. & King, J. F.
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