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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

Influence of a simulated HTGR environment on the mechanical properties of a commercial Ni-Cr-Mo-Fe alloy (Hastelloy Alloy X)

Description: The influence of a simulated advanced-reactor helium environment, containing 500 ..mu..atm H/sub 2//50 ..mu..atm CH/sub 4//50 ..mu..atm CO/approx. 1 ..mu..atm H/sub 2/O, on the mechanical properties of two heats of Hastelloy Alloy X is discussed. Simultaneous exposures in air and controlled-impurity helium at temperatures in the range of 650/sup 0/ to 1000/sup 0/C for times of 3000 h or more were performed. A combination of tensile testing, Charpy V-notch impact toughness testing, and creep testing was used to study the effects of reactor helium/metal interactions on the mechanical behavior of this alloy. Carburization was identified as the primary corrosion phenomenon. Increasing exposure time and temperature were observed to increase the depth of carburization. The increase in carbon concentration in the carburized zone suppressed the additional formation of M/sub 6/C, which is observed in air-aged specimens, and resulted in the precipitation of M/sub 23/C/sub 6/, a chromium-rich carbide variant. The precipitation of M/sub 23/C/sub 6/ in the carburized zone occurred primarily along grain and twin boundaries; however, matrix precipitation was also observed, the degree of which depended on exposure temperature. Strength and impact toughness properties were found to be controlled primarily by thermal aging reactions, with only a small effect related to the carburization. Although tensile and creep ductilities were decreased as a result of carburization, substantial ductility remained. Variation was observed between the two heats, the finer-grained heat appearing to be weaker in the high-temperature creep tests and also possibly more susceptible to a loss of creep strength as a result of carburization.
Date: December 1, 1979
Creator: Li, C.C.; Johnson, W.R. & Thompson, L.D.
Partner: UNT Libraries Government Documents Department

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

Effects of methane concentration on the controlled-impurity helium corrosion behavior of selected HTGR structural materials

Description: The corrosion behavior of candidate structural alloys in a series of three simulated advanced gas-cooled reactor environments at 900/sup 0/C (1652/sup 0/F), with methane concentration varied, is discussed. The alloys investigated include three wrought alloys, Hastelloy X, Inconel 617, and Incoloy 800H; two cast superalloys, Rene 100 and IN 713; one centrifugally cast alloy, HK 40; and an oxide-dispersion-strengthened alloy, MA 754. Corrosion behavior was found to be strongly dependent upon both the alloy chemistry and the environment. Oxidation, carburization, and/or mixed behavior was observed depending upon the specific conditions. An equilibrium thermodynamics approach has been used to predict alloy behavior and explain observations relevant to the understanding of gas/metal interactions in reactor helium, which inherently contains small amounts of reactive impurity species. Carburization was identified as the primary corrosion phenomenon of concern, and detailed analyses were performed to determine the susceptibility and control of carburization reactions. The presence of alumina scales, containing small amounts of titanium, was found to be particularly effective in inhibiting carburization. Small variations in methane concentration have been shown to have a dramatic effect upon the oxidation potential and subsequent corrosion behavior of the alloy systems.
Date: December 1, 1979
Creator: Johnson, W.R. & Thompson, L.D.
Partner: UNT Libraries Government Documents Department

Carburization of austenitic alloys by gaseous impurities in helium

Description: The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O and CO/sub 2/ was studied. Corrosion tests were conducted in a temperature range from 649 to 1000/sup 0/C (1200 to 1832/sup 0/F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 ..mu..atm H/sub 2/, 450 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 50 ..mu..atm H/sub 2/O for Environment A; 200 ..mu..atm H/sub 2/, 100 ..mu..atm CO, 20 ..mu..atm CH/sub 4/, 50 ..mu..atm H/sub 2/O and 5 ..mu..atm CO/sub 2/ for Environment B; 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and < 0.5 ..mu..atm H/sub 2/O for Environment C; and 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 1.5 ..mu..atm H/sub 2/O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys.
Date: March 1980
Creator: Lai, G. Y. & Johnson, W .R.
Partner: UNT Libraries Government Documents Department

Compatibility of aluminide-coated Hastelloy x and Inconel 617 in a simulated gas-cooled reactor environment

Description: Commercially prepared aluminide coatings on Hastelloy X and Inconel 617 substrates were exposed to controlled-impurity helium at 850/sup 0/ and 950/sup 0/C for 3000 h. Optical and scanning electron (SEM) microscopy, electron microprobe profiles, and SEM X-ray mapping were used to evaluate and compare exposed and unexposed control samples. Four coatings were evaluated: aluminide, aluminide with platinum, aluminide with chromium, and aluminide with rhodium. With extended time at elevated temperature, nickel diffused into the aluminide coatings to form epsilon-phase (Ni/sub 3/Al). This diffusion was the primary cause of porosity formation at the aluminide/alloy interface.
Date: March 1, 1982
Creator: Chin, J.; Johnson, W. R. & Chen, K.
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

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

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

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

Description: Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.
Date: August 1, 1981
Partner: UNT Libraries Government Documents Department