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Improved Austenitic Steels for Power Plant Applications

Description: Using alloy design principles, an austenitic alloy, with base composition of Fe-16Cr-16Ni-2Mn-1Mo (in weight percent, wt%), was formulated to which up to 5 wt% Si and/or Al were added specifically to improve the oxidation resistance. Cyclic oxidation tests were carried out in air at 700 and 800 C for 1000 hours. For comparison, Fe-18Cr-8Ni type-304 stainless steel alloys was also tested. The results showed that at 700 C, all the alloys were twice as oxidation resistant as the type-304 alloy (i.e., the experimental alloys showed weight gains about half that of type-304). Surprisingly, at 800 C, alloys that contained both Al and Si additions were less oxidation resistant than the type-304 alloy. However, alloys containing only Si additions were significantly more oxidation resistant than the type 304 alloys (i.e., showed weight gains 4 times less than the type-304 alloy). Further, alloys with only Si additions pre-oxidized at 800 C, showed zero weight gain in subsequent testing for 1000 hours at 700 C. This implies the potential for producing in-situ protective coating for these alloys. Preliminary exposure tests (1%H2S at 700 C for 360 hrs) indicated that the Si-modified alloys are more sulfidation resistant than type-304 alloy. The mechanical properties of the alloys, modified with carbide forming elements, were also evaluated; and at 600, 700 and 800 C the yield stresses of the carbide modified alloys were twice that of type-304 stainless steel. In this temperature range, the tensile properties of these alloys were comparable to literature values for type-347 stainless steel. It should be emphasized that the microstructures of the carbide forming alloys were not optimized with respect to grain size, carbide size and/or carbide distribution. Also, presented are initial results of vari-strain weld tests used to determine parameters for joining these alloys.
Date: August 6, 2002
Creator: Alman, David E.; Dunning, John S.; Schrems, Karol K.; Rawers, James C.; Wilson, Rick D.; Hawk, Jeffrey A. et al.
Partner: UNT Libraries Government Documents Department

Unlimited Damage Accumulation in Metallic Materials Under Cascade-Damage Conditions

Description: Most experiments on neutron or heavy-ion cascade-produced irradiation of pure metals and metallic alloys demonstrate unlimited void growth as well as development of the dislocation structure. In contrast, the theory of radiation damage predicts saturation of void swelling at sufficiently high irradiation doses and, accordingly, termination of accumulation of interstitial-type defects. It is shown in the present paper that, under conditions of steady production of one-dimensionally (1-D) mobile clusters of self-interstitial atoms (SIAs) in displacement cascades, any one of the following three conditions can result in indefinite damage accumulation. First, if the fraction of SIAs generated in the clustered form is smaller than some finite value of the order of the dislocation bias factor. Second, if solute, impurity or transmuted atoms form atmospheres around voids and repel the SIA clusters. Third, if spatial correlations between voids and other defects, such as second-phase precipitates and dislocations, exist that provide shadowing of voids from the SIA clusters. The driving force for the development of such correlations is the same as for void lattice formation and is argued to be always present under cascade-damage conditions. It is emphasised that the mean-free path of 1-D migrating SIA clusters is typically at least an order of magnitude longer than the average distance between microstructural defects; hence spatial correlations on the same scale should be taken into consideration. A way of developing a predictive theory is discussed. An interpretation
Date: September 1, 2008
Creator: Barashev, Aleksandr & Golubov, Stanislav I
Partner: UNT Libraries Government Documents Department

Analysis of nanometer-scale precipitation in a rapidly solidified stainless steel

Description: The authors have rapid-solidification-processed many stainless steels by gas atomization and achieved strength improvements of over 50% relative to conventionally-processed stainless steels with concomitant improvement in corrosion and oxidation behavior. These strength improvements are most pronounced after aging treatments when elevated concentrations of oxygen and vanadium are present in the stainless steel. An austenitic (FCC) stainless steel was prepared by gas atomization and consolidated by hot extrusion at 900 C. These specimens were heat treated for 1 hour at 1,000 C and aged at 600 C for 500 hours. The microstructure of each alloy composition was observed in TEM with bright field imaging. After aging, most alloys showed the same precipitate morphology as before aging. An obvious change, however, was found only in the alloy with highest oxygen content. A high number density of 15 to 20 nm diameter precipitates was measured in this alloy. Moreover, with weak-beam dark field imaging, a very high number density of coherent, 6 to 10 nm diameter precipitates is observed throughout the matrix by Moire fringe contrast. An atom probe field ion microscopy (APFIM) investigation showed that FIM provides high contrast imaging the precipitates. In order to get a more global view of the structure, energy-filtered composition imaging on a LEO EM 912 was used to map the oxygen and nitrogen in carbon extraction replicas of the aged specimens. These images confirm that the 18 nm precipitates are oxides, however, it appears that the 8 nm precipitates are not extracted.
Date: March 21, 1997
Creator: Wisutmethangoon, S.; Kelly, T.F.; Camus, P.P.; Flinn, J.E.; Larson, D.J. & Miller, M.K.
Partner: UNT Libraries Government Documents Department

Proceedings of the Eight Annual Conference on Fossil Energy Materials

Description: Objective of the meeting was to conduct R and D on materials for longer-term fossil energy applications as well as for generic needs of various fossil fuel technologies. The work is divided into ceramics, new alloys, corrosion, and technology assessment/transfer. The 39 papers are arranged under the session headings: ceramics, ceramics and new alloys, and intermetallics and advanced austenitics; a workshop on new materials development and applications is summarized briefly. The papers are processed separately for the data base.
Date: August 1, 1994
Creator: Cole, N. C. & Judkins, R. R.
Partner: UNT Libraries Government Documents Department

ITER breeding blanket design

Description: A breeding blanket design has been developed for ITER to provide the necessary tritium fuel to achieve the technical objectives of the Enhanced Performance Phase. It uses a ceramic breeder and water coolant for compatibility with the ITER machine design of the Basic Performance Phase. Lithium zirconate and lithium oxide am the selected ceramic breeders based on the current data base. Enriched lithium and beryllium neutron multiplier are used for both breeders. Both forms of beryllium material, blocks and pebbles are used at different blanket locations based on thermo-mechanical considerations and beryllium thickness requirements. Type 316LN austenitic steel is used as structural material similar to the shielding blanket. Design issues and required R&D data are identified during the development of the design.
Date: December 31, 1995
Creator: Gohar, Y.; Cardella, A.; Ioki, K.; Lousteau, D.; Mohri, K.; Raffray, R. et al.
Partner: UNT Libraries Government Documents Department

The Tritium Performance of Alloy 22-13-5

Description: Previously published studies of the performance of high strength austenitic steels and superalloys in tritium have demonstrated significant shortcomings in toughness and sensitivity to decay helium in the metal matrix. The alloy 22Cr-13Ni-5Mn exhibits high cracking thresholds in hydrogen, and promising performance in tritium-charged and aged smooth tensile specimens. It is readily forged to strengths beyond 690 MPa, and is commercially available. The tensile performance of 22-13-5 is compared to that of 21-6-9 and JBK-75. Aspects of a development program are outlined.
Date: June 1, 2001
Creator: Robinson, Steven L.
Partner: UNT Libraries Government Documents Department

Advanced nondestructive examination technologies for measuring fatigue damage in nuclear power plant components

Description: This paper presents recent results from an ongoing project at the Idaho National Engineering Laboratory (INEL) to develop advanced nondestructive methods to characterize the aging degradation of nuclear power plant pressure boundary components. One of the advanced methods, positron annihilation, is being developed for in situ characterization of fatigue damage in nuclear power plant piping and other components. This technique can detect and correlate the microstructural changes that are precursors of fatigue cracking in austenitic stainless steel components. In fact, the initial INEL test results show that the method can detect fatigue damage in stainless steel ranging from a few percent of the fatigue life up to 40 percent.
Date: December 1, 1995
Creator: MacDonald, P.E.; Shah, V.N. & Akers, D.W.
Partner: UNT Libraries Government Documents Department

Atomistic simulation of the hydrogen-induced fracture process in an iron-based superalloy

Description: Austenitic superalloys exhibit dramatic reductions in ductility and crack growth resistance when high fugacity hydrogen and hydrogen-producing environments trigger a change in fracture mode from microvoid coalescence to slip band and intergranular fracture. Of particular importance is the change to intergranular fracture. We have therefore combined the Embedded Atom Method (EAM) with Monte Carlo simulations and molecular dynamics calculations to help define the effects of hydrogen on segregation and fracture at the atomic level. Nickel was used to simulate the face-centered-cubic austenite lattice while symmetric and asymmetric {sigma}9 tilt boundaries were used to simulate grain boundaries. These simulations show that grain boundaries are strong trap sites for hydrogen. They further show that hydrogen dramatically reduces the bond strength between atoms at grain boundary sites while inhibiting dislocation generation.
Date: December 31, 1995
Creator: Moody, N.R.; Foiles, S.M.; Baskes, M.I. & Angelo, J.E.
Partner: UNT Libraries Government Documents Department

A guide for the ASME code for austenitic stainless steel containment vessels for high-level radioactive materials

Description: The design and fabrication criteria recommended by the US Department of Energy (DOE) for high-level radioactive materials containment vessels used in packaging is found in Section III, Division 1, Subsection NB of the ASME Boiler and Pressure Vessel Code. This Code provides material, design, fabrication, examination, and testing specifications for nuclear power plant components. However, many of the requirements listed in the Code are not applicable to containment vessels made from austenitic stainless steel with austenitic or ferritic steel bolting. Most packaging designers, engineers, and fabricators are intimidated by the sheer volume of requirements contained in the Code; consequently, the Code is not always followed and many requirements that do apply are often overlooked during preparation of the Safety Analysis Report for Packaging (SARP) that constitutes the basis to evaluate the packaging for certification.
Date: June 1, 1995
Creator: Raske, D.T.
Partner: UNT Libraries Government Documents Department

High resolution interface nanochemistry and structure. Progress report, December 1, 1994--November 30, 1995

Description: Progress is described in the following research areas concerned with high resolution interface nanochemistry and structure: ceramic interfaces and grain boundaries; metal/alpha (6H)-SiC(0001) interfaces; oxygen distributions in monolithic silicon carbide; SiC/nitride and metal on nitride interfaces; and interface synthesis.
Date: August 15, 1995
Creator: Carpenter, R.W.
Partner: UNT Libraries Government Documents Department

Phase transformations and microstructure development in low alloy steel welds

Description: Microstructure development in low alloy steel welds depends on various phase transformations that are a function of weld heating and cooling. The phase changes include non-metallic oxide inclusion formation in the liquid state, weld pool solidification, and solid state transformations. In this paper the mechanism of inclusion formation during low alloy steel welding is considered and the model predictions are compared with published results. The effect of inclusions on the austenite to ferrite transformation kinetics is measured and the mechanisms of transformation are discussed. The austenite gain development is related to the driving force for transformation of {delta} ferrite to austenite.
Date: July 1, 1995
Creator: Babu, S.S.; David, S.A. & Vitek, J.M.
Partner: UNT Libraries Government Documents Department

Direct Observation of Phase Transformations in Austenitic Stainless Steel Welds Using In-situ Spatially Resolved and Time-resolved X-ray Diffraction

Description: Spatially resolved x-ray diffraction (SRXRD) and time resolved x-ray diffraction (TRXRD) were used to investigate real time solid state phase transformations and solidification in AISI type 304 stainless steel gas tungsten arc (GTA) welds. These experiments were conducted at Stanford Synchrotron Radiation Laboratory (SSRL) using a high flux beam line. Spatially resolved observations of {gamma} {leftrightarrow} {delta} solid state phase transformations were performed in the heat affected zone (HAZ) of moving welds and time-resolved observations of the solidification sequence were performed in the fusion zone (FZ) of stationary welds after the arc had been terminated. Results of the moving weld experiments showed that the kinetics of the {gamma}{yields}{delta} phase transformation on heating in the HAZ were sufficiently rapid to transform a narrow region surrounding the liquid weld pool to the {delta} ferrite phase. Results of the stationary weld experiments showed, for the first time, that solidification can occur directly to the {delta} ferrite phase, which persisted as a single phase for 0.5s. Upon solidification to {delta}, the {delta} {yields} {gamma} phase transformation followed and completed in 0.2s as the weld cooled further to room temperature.
Date: September 23, 1999
Creator: Elmer, J.; Wong, J. & Ressler, T.
Partner: UNT Libraries Government Documents Department

Transformation plasticity in ductile solids. Annual progress report, June 1, 1993--May 31, 1994

Description: Research has addressed the role of martensitic transformation plasticity in the enhancement of toughness in high-strength austenitic steels, and the enhancement of formability in multiphase low-alloy sheet steels. In the austenitic steels, optimal processing has achieved a significant increase in strength level, in order to investigate the interaction of strain-induced transformation with the microvoid nucleation and shear localization mechanisms operating at ultrahigh strength levels. The degree of transformation interaction is sensitive to both strength level and degree of constraint. The stress-state dependence of transformation and fracture mechanisms has been investigated in model alloys, comparing behavior in uniaxial tension and blunt-notch tension specimens. A reformulated numerical constitutive model for transformation plasticity has allowed a more thorough analysis of transformation/fracture interactions, including local processes of microvoid nucleation. Processing of a new low alloy steel composition has been optimized to stabilize retained austenite by isothermal bainitic transformation after intercritical annealing. Results show a good correlation of uniform ductility with the austenite amount and stability, and new compositions are designed for improved stability.
Date: April 1, 1994
Creator: Olson, G. B.
Partner: UNT Libraries Government Documents Department

Fatigue of ferritic and austenitic steels. Final technical report, June 1, 1984--July 31, 1991

Description: One aim of this research program has been to increase our understanding of the mechanisms of fatigue crack growth as influenced by such parameters as the specific material, crack length, R-ratio, overloads, temperature and the environment. A second objective has been to utilize this understanding in the development of semi-empirical quantitative methods for the prediction of fatigue crack growth behavior. Although there are still many questions remaining concerning fatigue crack growth, we feel that some significant progress in meeting these two objectives has been made.
Date: December 31, 1991
Creator: McEvily, A. J.
Partner: UNT Libraries Government Documents Department

Proceedings of the Seventh Annual Conference on Fossil Energy Materials. Fossil Energy AR and TD Materials Program

Description: Objective of the AR&TD Materials Program is to conduct research and development on materials for longer-term fossil energy applications as well as for generic needs of various fossil fuel technologies. The 37 papers are arranged into 3 sessions: ceramics, new alloys/intermetallics, and new alloys/advanced austenitics. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.
Date: July 1, 1993
Creator: Cole, N. C. & Judkins, R. R.
Partner: UNT Libraries Government Documents Department

High Temperature and Pressure Steam-H2 Interaction with Candidate Advanced LWR Fuel Claddings

Description: This report summarizes the work completed to evaluate cladding materials that could serve as improvements to Zircaloy in terms of accident tolerance. This testing involved oxidation resistance to steam or H{sub 2}-50% steam environments at 800-1350 C at 1-20 bar for short times. A selection of conventional alloys, SiC-based ceramics and model alloys were used to explore a wide range of materials options and provide guidance for future materials development work. Typically, the SiC-based ceramic materials, alumina-forming alloys and Fe-Cr alloys with {ge}25% Cr showed the best potential for oxidation resistance at {ge}1200 C. At 1350 C, FeCrAl alloys and SiC remained oxidation resistant in steam. Conventional austenitic steels do not have sufficient oxidation resistance with only {approx}18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application and high Cr contents raise concern about {alpha}{prime} formation. Higher pressures (up to 20.7 bar) and H{sub 2} additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed in a H{sub 2}-50%H{sub 2}O environment at 10.3 bar. As some of the results regarding low-alloyed FeCrAl and Fe-Cr alloys were unexpected, further work is needed to fundamentally understand the minimum Cr and Al alloy contents needed for protective behavior in these environments in order to assist in alloy selection and guide alloy development.
Date: August 1, 2012
Creator: Pint, Bruce A
Partner: UNT Libraries Government Documents Department

Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

Description: Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be related to the formation of second-phases under irradiation, although further examination is required
Date: April 1, 2011
Creator: Busby, Jeremy T & Gussev, Maxim N
Partner: UNT Libraries Government Documents Department

Manufacture of Alumina-Forming Austenitic Steel Alloys by Conventional Casting and Hot-Working Methods

Description: Oak Ridge National Laboratory (ORNL) and Carpenter Technology Corporation (CarTech) participated in an in-kind cost share cooperative research and development agreement (CRADA) effort under the auspices of the Energy Efficiency and Renewable Energy (EERE) Technology Maturation Program to explore the feasibility for scale up of developmental ORNL alumina-forming austenitic (AFA) stainless steels by conventional casting and rolling techniques. CarTech successfully vacuum melted 301b heats of four AFA alloy compositions in the range of Fe-(20-25)Ni-(12-14)Cr-(3-4)Al-(l-2.5)Nb wt.% base. Conventional hot/cold rolling was used to produce 0.5-inch thick plate and 0.1-inch thick sheet product. ORNL subsequently successfully rolled the 0.1-inch sheet to 4 mil thick foil. Long-term oxidation studies of the plate form material were initiated at 650, 700, and 800 C in air with 10 volume percent water vapor. Preliminary results indicated that the alloys exhibit comparable (good) oxidation resistance to ORNL laboratory scale AFA alloy arc casting previously evaluated. The sheet and foil material will be used in ongoing evaluation efforts for oxidation and creep resistance under related CRADAs with two gas turbine engine manufacturers. This work will be directed to evaluation of AFA alloys for use in gas turbine recuperators to permit higher-temperature operating conditions for improved efficiencies and reduced environmental emissions. AFA alloy properties to date have been obtained from small laboratory scale arc-castings made at ORNL. The goal of the ORNL-CarTech CRADA was to establish the viability for producing plate, sheet and foil of the AFA alloys by conventional casting and hot working approaches as a first step towards scale up and commercialization of the AFA alloys. The AFA alloy produced under this effort will then be evaluated in related CRADAs with two gas turbine engine manufacturers for gas turbine recuperator applications.
Date: March 10, 2009
Creator: Brady, M. P.; Yamamoto, Y. & Magee, J. H.
Partner: UNT Libraries Government Documents Department

Report on thermal aging effects on tensile properties of ferritic-martensitic steels.

Description: This report provides an update on the evaluation of thermal-aging induced degradation of tensile properties of advanced ferritic-martensitic steels. The report is the first deliverable (level 3) in FY11 (M3A11AN04030103), under the Work Package A-11AN040301, 'Advanced Alloy Testing' performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing tensile data on aged alloys and a mechanistic model, validated by experiments, with a predictive capability on long-term performance. The scope of work is to evaluate the effect of thermal aging on the tensile properties of advanced alloys such as ferritic-martensitic steels, mod.9Cr-1Mo, NF616, and advanced austenitic stainless steel, HT-UPS. The aging experiments have been conducted over a temperature of 550-750 C for various time periods to simulate the microstructural changes in the alloys as a function of time at temperature. In addition, a mechanistic model based on thermodynamics and kinetics has been used to address the changes in microstructure of the alloys as a function of time and temperature, which is developed in the companion work package at ANL. The focus of this project is advanced alloy testing and understanding the effects of long-term thermal aging on the tensile properties. Advanced materials examined in this project include ferritic-martensitic steels mod.9Cr-1Mo and NF616, and austenitic steel, HT-UPS. The report summarizes the tensile testing results of thermally-aged mod.9Cr-1Mo, NF616 H1 and NF616 H2 ferritic-martensitic steels. NF616 H1 and NF616 H2 experienced different thermal-mechanical treatments before thermal aging experiments. NF616 H1 was normalized and tempered, and NF616 H2 was normalized and tempered and cold-rolled. By examining these two heats, we evaluated the effects of thermal-mechanical treatments on material microstructures and associated mechanical properties during long-term aging at elevated temperatures. Thermal aging experiments at different temperatures and ...
Date: May 10, 2012
Creator: Li, M.; Soppet, W.K.; Rink, D.L.; Listwan, J.T. & Natesan, K. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Irradiation creep at temperatures of 400 {degrees}C and below for application to near-term fusion devices

Description: To study irradiation creep at 400{degrees}C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400{degrees}C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330{degrees}C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys.
Date: December 31, 1996
Creator: Grossbeck, M.L.; Gibson, L.T. & Mansur, L.K.
Partner: UNT Libraries Government Documents Department

Advanced austenitic alloys for fossil power systems. CRADA final report

Description: In 1993, a Cooperative Research and Development Agreement (CRADA) was undertaken between Oak Ridge National Laboratory and ABB Combustion Engineering t examine advanced alloys for fossil power systems. Specifically, the use of advanced austenitic stainless steels for superheater/reheater construction in supercritical boilers was examined. The strength of cold-worked austenitic stainless steels was reviewed and compared to the strength and ductility of advanced austenitic stainless steels. The advanced stainless steels were found to retain their strength to very long times at temperatures where cold-worked standard grades of austenitic stainless steels became weak. Further, the steels exhibited better long-time stability than the stabilized 300 series stainless steels in either the annealed or cold worked conditions. Type 304H mill-annealed tubing was provided to ORNL for testing of base metal and butt welds. The tubing was found to fall within range of expected strength for 304H stainless steel. The composite 304/308 stainless steel was found to be stronger than typical for the weldment. Boiler tubing was removed from a commercial boiler for replacement by newer steels, but restraints imposed by the boiler owners did not permit the installation of the advanced steels, so a standard 32 stainless steel was used as a replacement. The T91 removed from the boiler was characterized.
Date: August 1, 1998
Creator: Swindeman, R.W.; Cole, N.C.; Canonico, D.A. & Henry, J.F.
Partner: UNT Libraries Government Documents Department

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

High-pressure tritium equipment

Description: Some solutions to problems of compressing and containing tritium gas to 200 MPa at 700 K are discussed The principal emphasis is on commercial compressors and high-pressure equipment that can be modified easily by the researcher for safe use with tritium. Experience with metal belows and diaphragm compressors has been favorable. Selection of materials, fittings and gauges for high- pressure tritium work also is reviewed briefly.
Date: December 31, 1976
Creator: Coffin, D.O.
Partner: UNT Libraries Government Documents Department