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Development of a Controlled Material Specification for Alloy 617 for Nuclear Applications

Description: Investigation is conducted in an effort to refine the standard specifications of Alloy 617 for the Very High Temperature Reactor applications. Background, motivation and rationale of the investigation are discussed. Historical data generated from various heats of the alloy are collected, sorted, and analyzed. The analyses include examination of mechanical property data and corresponding heat chemical composition, discussion on previous Alloy 617 specification development effort at the Oak Ridge National Laboratory, and assessment of the strengthening elements and mechanisms of the alloy. Based on the analyses, literature review, and knowledge of Ni base alloys, a tentative refined specification is recommended. Future work for verifying and improving the tentative refined specification is also suggested.
Date: May 1, 2005
Creator: Ren, Weiju
Partner: UNT Libraries Government Documents Department

Compatibility of gas turbine materials with steam cooling

Description: Objective is to investigate performance of gas turbine materials in steam environment and evaluate remedial measures for alleviating the severity of the problem. Three superalloys commonly used in gas turbines were exposed to 3 steam environments containing different impurity levels for 2 to 6 months. Results: Cr2O3-forming alloys containing 1-4% Al such as IN 738 are susceptible to heavy internal oxidation of Al. High Al (>5%) alloys in which continuous Al2O3 scale can be formed, may not be susceptible to such attack. Deposition of salts from steam will accentuate hot corrosion problems. Alloys with higher Cr content such as X-45 are generally less prone to hot corrosion. The greater damage observed in IN 617 make this alloy less attractive for gas turbines with steam cooling. Electrochemical impedance spectroscopy is a good nondestructive method to evaluate microstructural damage.
Date: December 31, 1995
Creator: Desai, V.; Tamboli, D. & Patel, Y.
Partner: UNT Libraries Government Documents Department

Controlled Chemistry Helium High Temperature Materials Test Loop

Description: A system to test aging and environmental effects in flowing helium with impurity content representative of the Next Generation Nuclear Plant (NGNP) has been designed and assembled. The system will be used to expose microstructure analysis coupons and mechanical test specimens for up to 5,000 hours in helium containing potentially oxidizing or carburizing impurities controlled to parts per million levels. Impurity levels in the flowing helium are controlled through a feedback mechanism based on gas chromatography measurements of the gas chemistry at the inlet and exit from a high temperature retort containing the test materials. Initial testing will focus on determining the nature and extent of combined aging and environmental effects on microstructure and elevated temperature mechanical properties of alloys proposed for structural applications in the NGNP, including Inconel 617 and Haynes 230.
Date: August 1, 2005
Creator: WRight, Richard N.
Partner: UNT Libraries Government Documents Department

Ultra-Supercritical Steam Corrosion

Description: Efficiency increases in fossil energy boilers and steam turbines are being achieved by increasing the temperature and pressure at the turbine inlets well beyond the critical point of water. To allow these increases, advanced materials are needed that are able to withstand the higher temperatures and pressures in terms of strength, creep, and oxidation resistance. As part of a larger collaborative effort, the Albany Research Center (ARC) is examining the steam-side oxidation behavior for ultrasupercritical (USC) steam turbine applications. Initial tests are being done on six alloys identified as candidates for USC steam boiler applications: ferritic alloy SAVE12, austenitic alloy Super 304H, the high Cr-high Ni alloy HR6W, and the nickel-base superalloys Inconel 617, Haynes 230, and Inconel 740. Each of these alloys has very high strength for its alloy type. Three types of experiments are planned: cyclic oxidation in air plus steam at atmospheric pressure, thermogravimetric ana lysis (TGA) in steam at atmospheric pressure, and exposure tests in supercritical steam up to 650 C (1202 F) and 34.5 MPa (5000 psi). The atmospheric pressure tests, combined with supercritical exposures at 13.8, 20.7, 24.6, and 34.5 MPa (2000, 3000, 4000, and 5000 psi) should allow the determination of the effect of pressure on the oxidation process.
Date: April 22, 2003
Creator: Holcomb, G. R.; Alman, D. E.; Bullard, S. B.; Covino, B. S., Jr.; Cramer, S. D. & Ziomek-Moroz, M.
Partner: UNT Libraries Government Documents Department

Tritium Permeability of Incoloy 800H and Inconel 617

Description: Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950 C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm) - three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.
Date: September 1, 2011
Creator: Winston, Philip; Calderoni, Pattrick & Humrickhouse, Paul
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan (PLN-2804)

Description: DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for application in heat exchangers ...
Date: April 1, 2008
Creator: Wright, J. K.
Partner: UNT Libraries Government Documents Department

Hydrogen Permeability of Incoloy 800H, Inconel 617, and Haynes 230 Alloys

Description: A potential issue in the design of the NGNP reactor and high-temperature components is the permeation of fission generated tritium and hydrogen product from downstream hydrogen generation through high-temperature components. Such permeation can result in the loss of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system. The issue will be addressed in the engineering design phase, and requires knowledge of permeation characteristics of the candidate alloys. Of three potential candidates for high-temperature components of the NGNP reactor design, the hydrogen permeability has been documented well only for Incoloy 800H, but at relatively high partial pressures of hydrogen. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. The hydrogen permeability of Haynes 230 has not been published. To support engineering design of the NGNP reactor components, the hydrogen permeability of Inconel 617 and Haynes 230 were determined using a measurement system designed and fabricated at the Idaho National Laboratory. The performance of the system was validated using Incoloy 800H as reference material, for which the permeability has been published in several journal articles. The permeability of Incoloy 800H, Inconel 617 and Haynes 230 was measured in the temperature range 650 to 950 °C and at hydrogen partial pressures of 10-3 and 10-2 atm, substantially lower pressures than used in the published reports. The measured hydrogen permeability of Incoloy 800H and Inconel 617 were in good agreement with published values obtained at higher partial pressures of hydrogen. The hydrogen permeability of Inconel 617 and Haynes 230 were similar, about 50% greater than for Incoloy 800H and with similar temperature dependence.
Date: July 1, 2010
Creator: Calderoni, Pattrick
Partner: UNT Libraries Government Documents Department

Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

Description: An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.
Date: December 1, 1983
Partner: UNT Libraries Government Documents Department

High-temperature low-cycle fatigue and tensile properties of Hastelloy X and alloy 617 in air and HTGR-helium

Description: Results of strain controlled fatigue and tensile tests are presented for two nickel base solution hardened alloys which are reference structural alloys for use in several high temperature gas cooled reactor concepts. These alloys, Hastelloy X Inconel 617, were tested at temperatures ranging from room temperature to 871/sup 0/C in air and impure helium. Materials were tested in the solution annealed as well as in the pre-aged condition where aging consisted of isothermal exposure at one of several temperatures for periods of up to 20,000 h. Comparisons are also given between the strain controlled fatigue lives of these alloys and several other commonly used alloys all tested at 538/sup 0/C.
Date: January 1, 1981
Creator: Strizak, J.P.; Brinkman, C.R. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

Corrosion performance of materials for advanced combustion systems

Description: Conceptual designs of advanced combustion systems that utilize coal as a feedstock require high-temperature furnaces and heat transfer surfaces capable of operating at more elevated temperatures than those prevalent in current coal-fired power plants. The combination of elevated temperatures and hostile combustion environments necessitates development/application of advanced ceramic materials in these designs. This report characterizes the chemistry of coal-fired combustion environments over the wide temperature range that is of interest in these systems and discusses preliminary experimental results on several materials (alumina, Hexoloy, SiC/SiC, SiC/Si{sub 3}N{sub 4}/Si{sub 3}N{sub 4}, ZIRCONIA, INCONEL 677 and 617) with potential for application in these systems.
Date: December 1, 1993
Creator: Natesan, K.; Yanez-Herrero, M. & Fornasieri, C.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Materials Research and Development Plan

Description: DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for application in heat exchangers ...
Date: September 1, 2010
Creator: Wright, J. K.
Partner: UNT Libraries Government Documents Department

Oxidation of selected alloys during 25,000 h in superheated steam at 482 and 538/sup 0/C

Description: The corrosion of several ferritic and austenitic materials in flowing superheated steam at 482 and 538/sup 0/C (900 and 1000/sup 0/F) were studied. Results obtained during the first 12,000 h of the test were presented previously. Results obtained during the first 25,000 h are summarized. The test specimens are mounted in a nonrecirculating loop that receives steam from the superheater circuit of a fossil-fired power plant. At both temperatures all materials exhibited parabolic oxidation kinetics during the first year and subsequently have oxidized at low constant rates. The ferritic steels containing 2 1/4 and 9% Cr have oxidized at about the same rates, averaging 4.2 and 8.6 ..mu..m/year (0.17 and 0.34 mils/year) at 482 and 538/sup 0/C, respectively, after the first year. Sandvik HT-9 (11.4% Cr) has corroded at slightly lower rates. Annealed and Cold-worked surfaces of these alloys have exhibited identical behavior. At 482/sup 0/C all materials have retained their corrosion products completely, but at 538/sup 0/C some began experiencing exfoliation after 12,000 h. Data suggest that a high silicon content in the alloy minimizes exfoliation. Cold-worked surfaces of alloy 800 are corroding at lower rates than annealed and pickled ones, but in both cases the rates are very low. Alloy 800 specimens that had been intergranularly corroded before exposure to steam are oxidizing at much higher rates, but intergranular penetration has not progressed. Type 304 stainless steel is corroding nonuniformly, but the attack rates are low at both temperatures. Alloy 617 is corroding at the lowest rate of any material in the loop; even after 25,000 h surface films are thin enough to show interference colors.
Date: March 1, 1980
Creator: Griess, J.C. & Maxwell, W.A.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

Description: This report presents the results of work performed from October 1, 1979 through December 31, 1979. 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 progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.
Date: April 18, 1980
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

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

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 1, 1980
Creator: Lai, G.Y. & Johnson, W.R.
Partner: UNT Libraries Government Documents Department

Materials research for the clean utilization of coal. Quarterly progress report, April-June 1981

Description: Effort this quarter has been concentrated on the book Construction Materials for Coal Conversion - Performance and Properties Data. The status of the various subsections of Section A (Materials Considerations and Performance Data) is: (1) Operating requirements - completed; (2) Performance Data and Candidate Materials - being drafted in final form. The assembling of test data for Section B is essentially complete and analysis of this data is in progress. Data was obtained on the creep of a fused cast ..cap alpha.. + ..beta.. alumina (Monofrax A) under thermal cycling conditions and on silicon nitride using both linear variable differential transformers and specimen dimension measurements.
Date: January 1, 1981
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

Long-term oxidation of selected alloys in superheated steam at 482 and 538/sup 0/C

Description: The oxidation of several Cr-Mo steels and austenitic materials in superheated steam at 482 and 538/sup 0/C (900 and 1000/sup 0/F) is studied. The investigation was conducted in a once-through loop that received steam from the superheater circuit of the Bartow Power Plant of the Florida Power Corporation. This report presents the results from this investigation, which was terminated after 28,339 h when the mode of power plant operation was changed from baseload to peaking.
Date: July 1, 1981
Creator: Griess, J.C. & Maxwell, W.A.
Partner: UNT Libraries Government Documents Department

Methods for very high temperature design

Description: Design rules and procedures for high-temperature, gas-cooled reactor components are being formulated as an ASME Boiler and Pressure Vessel Code Case. A draft of the Case, patterned after Code Case N-47, and limited to Inconel 617 and temperatures of 982/degree/C (1800/degree/F) or less, will be completed in 1989 for consideration by relevant Code committees. The purpose of this paper is to provide a synopsis of the significant differences between the draft Case and N-47, and to provide more complete accounts of the development of allowable stress and stress rupture values and the development of isochronous stress vs strain curves, in both of which Oak Ridge National Laboratory (ORNL) played a principal role. The isochronous curves, which represent average behavior for many heats of Inconel 617, were based in part on a unified constitutive model developed at ORNL. Details are also provided of this model of inelastic deformation behavior, which does not distinguish between rate-dependent plasticity and time-dependent creep, along with comparisons between calculated and observed results of tests conducted on a typical heat of Inconel 617 by the General Electric Company for the Department of Energy. 4 refs., 15 figs., 1 tab.
Date: January 1, 1989
Creator: Blass, J.J.; Corum, J.M. & Chang, S.J.
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

Corrosion studies in molten calcium chloride with chlorine

Description: This study is aimed at testing new materials for use in molten salt processing of plutonium. Because of the high corrosiveness of chlorine, present materials have a high rate of failure. Materials less subject to corrosion are needed to minimize costs resulting from rapid failure of sparge tubes, stirring apparatus, and crucibles; to reduce the quantity of plutonium-contaminated scrap; and to improve the purity of the plutonium product. The processing environment of molten CaCl{sub 2}--CaO salts, molten plutonium, and chlorine-oxygen gas at temperatures from 750{degree} to 900{degree} is extremely severe. Materials with resistance to both corrosion and mechanical failure are desired. Also, the incorporation of corrosion products into the final plutonium product cannot exceed the allowable impurity limits. We require materials for crucibles, sparge tubes, stirrers, and containment and pull cans. Four metallic and two ceramics materials were tested. The metallic materials were Inconel-601, Inconel-617, tantalum, and tungsten. Silicon nitride and magnesium oxide were the ceramics tested.
Date: January 1, 1990
Creator: McLaughlin, D.F. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Science and Technology Center); Sessions, C.E. & Marra, J.E. (Westinghouse Savannah River Co., Aiken, SC (USA))
Partner: UNT Libraries Government Documents Department