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Fireside Corrosion in Oxy-Fuel Combustion of Coal

Description: Oxy-fuel combustion is based on burning fossil fuels in a mixture of recirculated flue gas and oxygen, rather than in air. An optimized oxy-combustion power plant will have ultra-low emissions since the flue gas that results from oxy-fuel combustion consists almost entirely of CO2 and water vapor. Once the water vapor is condensed, it is relatively easy to sequester the CO2 so that it does not escape into the atmosphere. A variety of laboratory tests comparing air-firing to oxy-firing conditions, and tests examining specific simpler combinations of oxidants, were conducted at 650-700 C. Alloys studied included model Fe-Cr and Ni-Cr alloys, commercial ferritic steels, austenitic steels, and nickel base superalloys. The observed corrosion behavior shows accelerated corrosion even with sulfate additions that remain solid at the tested temperatures, encapsulation of ash components in outer iron oxide scales, and a differentiation between oxy-fuel combustion flue gas recirculation choices.
Date: August 1, 2012
Creator: Holcomb, Gordon R.; Tylczak, Joseph; Meier, G.H.; K., Jung.; Mu, N.; Yanar, N.M. et al.
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

Report on sodium compatibility of advanced structural materials.

Description: This report provides an update on the evaluation of sodium compatibility of advanced structural materials. The report is a deliverable (level 3) in FY11 (M3A11AN04030403), under the Work Package A-11AN040304, 'Sodium Compatibility of Advanced Structural Materials' performed by Argonne National Laboratory (ANL), as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing corrosion and tensile data from the standpoint of sodium compatibility of advanced structural alloys. The scope of work involves exposure of advanced structural alloys such as G92, mod.9Cr-1Mo (G91) ferritic-martensitic steels and HT-UPS austenitic stainless steels to a flowing sodium environment with controlled impurity concentrations. The exposed specimens are analyzed for their corrosion performance, microstructural changes, and tensile behavior. Previous reports examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design, fabrication, and construction of a forced convection sodium loop for sodium compatibility studies of advanced materials. This report presents the results on corrosion performance, microstructure, and tensile properties of advanced ferritic-martensitic and austenitic alloys exposed to liquid sodium at 550 C for up to 2700 h and at 650 C for up to 5064 h in the forced convection sodium loop. The oxygen content of sodium was controlled by the cold-trapping method to achieve {approx}1 wppm oxygen level. Four alloys were examined, G92 in the normalized and tempered condition (H1 G92), G92 in the cold-rolled condition (H2 G92), G91 in the normalized and tempered condition, and hot-rolled HT-UPS. G91 was included as a reference to compare with advanced alloy, G92. It was found that all four alloys showed weight loss after sodium exposures at 550 and 650 C. The weight loss of the four alloys was comparable after sodium exposures at ...
Date: July 9, 2012
Creator: Li, M.; Natesan, K.; Momozaki, Y.; Rink, D.L.; Soppet, W.K. & Listwan, J.T. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Fireside Corrosion in Oxy-fuel Combustion of Coal

Description: Oxy-fuel combustion is burning a fuel in oxygen rather than air. The low nitrogen flue gas that results is relatively easy to capture CO{sub 2} from for reuse or sequestration. Corrosion issues associated with the environment change (replacement of much of the N{sub 2} with CO{sub 2} and higher sulfur levels) from air- to oxy-firing were examined. Alloys studied included model Fe-Cr alloys and commercial ferritic steels, austenitic steels, and nickel base superalloys. The corrosion behavior is described in terms of corrosion rates, scale morphologies, and scale/ash interactions for the different environmental conditions.
Date: May 20, 2012
Creator: Holcomb, G. R.; Tylczak, J.; Meier, G. H.; Lutz, B.; Jung, K.; Mu, N. et al.
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

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

Technetium Waste Form Development - Progress Report

Description: Analytical electron microscopy using SEM and TEM has been used to analyze a ~5 g. ingot with composition 71.3 wt% 316SS-5.3 wt% Zr-13.2 wt% Mo-4.0 wt% Rh-6.2 wt% Re prepared at the Idaho National Laboratory. Four phase fields have been identified two of which are lamellar eutectics, with a fifth possibly present. A Zr rich phase was found distributed as fine precipitate, ~10µm in diameter, often coating large cavities. A Mo-Fe-Re-Cr lamellar eutectic phase field appears as blocky regions ~30µm in diameter, surrounded by a Fe-Mo-Cr lamellar eutectic phase field, and that in turn is surrounded by a Zr-Fe-Rh-Mo-Ni phase field. The eutectic phase separation reactions are different. The Mo-Fe-Re-Cr lamellar eutectic appears a result of austenitic steel forming at lower volume fraction within an Mo-Fe-Re intermetallic phase, whereas the Fe-Mo-Cr lamellar eutectic may be a result of the same intermetallic phase forming within a ferritic steel phase. Cavitation may have arisen either as a result of bubbles, or from loss of equiaxed particles during specimen preparation.
Date: January 7, 2009
Creator: Gelles, David S.; Ermi, Ruby M.; Buck, Edgar C.; Seffens, Rob J. & Chamberlin, Clyde E.
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

Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

Description: This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.
Date: November 17, 2004
Creator: Balachov, Iouri I.; Kobayashi, Takao; Tanzella, Francis; Jayaweera, Indira; Jayaweera, Palitha; Kinnunen, Petri et al.
Partner: UNT Libraries Government Documents Department

Aluminide Coatings for Power-Generation Applications

Description: Aluminide coatings are of interest for many high temperature applications because of the possibility of improving the oxidation of structural alloys by forming a protective external alumina scale. In order to develop a comprehensive lifetime evaluation approach for aluminide coatings used in fossil energy systems, some of the important issues have been addressed in this report for aluminide coatings on Fe-based alloys (Task I) and on Ni-based alloys (Task II). In Task I, the oxidation behavior of iron aluminide coatings synthesized by chemical vapor deposition (CVD) was studied in air + 10vol.% H{sub 2}O in the temperature range of 700-800 C and the interdiffusion behavior between the coating and substrate was investigated in air at 500-800 C. Commercial ferritic (Fe-9Cr-1Mo) and type 304L (Fe-18Cr-9Ni, nominally) austenitic stainless steels were used as the substrates. For the oxidation study, the as-deposited coating consisted of a thin (<5 {micro}m), Al-rich outer layer above a thicker (30-50 {micro}m), lower Al inner layer. The specimens were cycled to 1000 1-h cycles at 700 C and 500 1-h cycles at 800 C, respectively. The CVD coating specimens showed excellent performance in the water vapor environment at both temperatures, while the uncoated alloys were severely attacked. These results suggest that an aluminide coating can substantially improve resistance to water vapor attack under these conditions. For the interdiffusion study, the ferritic and austenitic steels were coated with relatively thicker aluminide coatings consisting of a 20-25 {micro}m outer layer and a 150-250 {micro}m inner layer. The composition profiles before and after interdiffusion testing (up to 5,000h) were measured by electron probe microanalysis (EPMA). The decrease of the Al content at the coating surface was not significant after extended diffusion times ({le} 5,000h) at temperatures {le} 700 C. More interdiffusion occurred at 800 C in coatings on both Fe- 9Cr-1Mo ...
Date: November 17, 2003
Creator: Zhang, Y
Partner: UNT Libraries Government Documents Department

Boron Addition to Model Austenitic Steels and void Nucleation

Description: Fe-15Cr-16Ni, -0.25Ti, -500appmB, and -0.25Ti-500appmB have been irradiated in FFTF/MOTA over a wide range of dose rate which covers more than two orders difference in magnitude, within the very limited temperature range of 387-444 C. The effects of dose rate and boron addition on swelling are examined. Lower dose rates increase the swelling by shortening the incubation dose for swelling. Addition of boron does not significantly change the swelling nor the dose rate dependence of swelling for both the ternary and Ti-modified alloy. The helium pressure of cavities is found to be much smaller than the surface tension at every irradiation condition including the lowest dose and dose rate, helium generated by boron transmutant does not play any role in cavity formation in this experiment. Cavities form without helium. The difference in cavity morphology by boron addition is most likely caused by formation of borides and by lithium.
Date: October 30, 2003
Creator: Okita, T; Wolfer, W G; Garner, F A & Sekimura, N
Partner: UNT Libraries Government Documents Department

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

Lead-bismuth target design for the subcritical multiplier (SCM) of the accelerator driven test facility (ADTF).

Description: A lead-bismuth eutectic (LBE) target design concept has been developed to drive the subcritical multiplier (SCM) of the accelerator-driven test facility (ADTF). This report gives the target design description, the results from the parametric studies, and the design analyses including physics, heat-transfer, hydraulics, structural, radiological, and safety analyses. The design is based on a coaxial geometrical configuration to minimize the target footprint and to maximize the utilization of the spallation neutrons. The target is installed vertically along the SCM axis. LBE is the target material and the target coolant. Ferritic steel (HT-9 alloy) is the selected structural material based on the current database and the design analyses. Austenitic steel (Type 316 stainless steel) is the backup choice. A uniform proton beam is employed to perform the spallation process. The proton beam has 8.33-mA current and 8.14-cm radius resulting in a current density of 40 {micro}A/cm{sup 2}. The beam power is 5 MW and the proton energy is 600 MeV. The beam tube has 10-cm radius to accommodate the halo current. A hemi-spherical geometry is used for the target window, which is connected to the beam tube. A conical target window with a rounded tip is also considered since it has a lower average temperature relative to the spherical geometry. The beam tube is enclosed inside two coaxial tubes to provide inlet and outlet manifolds for the LBE coolant. The inlet and the outlet coolant manifolds and the proton beam are entered from the top above the SCM. Several design constraints are developed and utilized for the target design process to satisfy different engineering requirements and to minimize the design development time and cost.
Date: April 18, 2002
Creator: Gohar, Y.; Finck, P. J.; Krajtl, L.; Herceg, J.; Pointer, W. D.; Saiveau, J. 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

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

Adsorption of Pu(IV) Polymer onto 304L Stainless Steel

Description: 'The report, Technical Basis for Safe Operations with Pu-239 Polymer in NMS S Operating Facilities (F H Areas), (WSRC-TR-99-00008) was issued in an effort to upgrade the Authorization Basis (AB) for H Area facilities relative to nuclear criticality. At the time, insufficient data were found in the literature to quantify the adsorption of Pu polymer onto the surfaces of stainless steel tanks. Additional experimental or literature information on the adsorption of Pu(IV) polymer and its removal was deemed necessary to support the H Area AB. The results obtained are also applicable to processing in F Area facilities.Additional literature sources suggest that adsorption on the tank walls should not be a safety concern. The sources show that the amount of Pu polymer that adsorbs from a solution comes to a limiting amount in 5 to 7 days after which no additional Pu is adsorbed. Adsorption increases with Pu concentration and decreases with acid concentration. The adsorbed amounts are small varying from 0.5 mg/cm2 for a 0.5 g/l Pu / 0.5M HNO3 solution to 11 mg/cm2 for a 1-3 g/l Pu / 0.1M HNO3 solution. Additionally, acid concentrations greater than 0.1M will remove a percentage of adsorbed Pu.The experimental results have generally confirmed much of what has been reported in the literature. Specifically, adsorption onto stainless steel was found to increase with increased Pu concentration, and decreased acid concentration. The amount adsorbed was found to come to a limiting amount after 5 to 7 days. Pu adsorbed as polymer was found to be harder to remove than if it was adsorbed as Pu(IV). The amount of Pu adsorbed as polymer was found to be almost an order of magnitude more than that from a similar concentration Pu(IV) solution. Unlike the literature, only a slight increase in adsorption values was found when the ...
Date: July 16, 1999
Creator: Bronikowski, M.G.
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

Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

Description: A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 {approximately} 325 C, up to {approximately}10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels.
Date: November 30, 1998
Creator: Gan, J.; Stoller, R.E. & Was, G.S.
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

Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

Description: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.
Date: April 1998
Creator: Chopra, O. K.; Chung, H. M. & Gruber, E. E.
Partner: UNT Libraries Government Documents Department

Irradiation-induced instability of MnS precipitates and its possible contribution to IASCC in light water reactors

Description: Although a number of candidate mechanisms have been proposed to participate in the IASCC phenomenon, it is not clear at this time that all of the contributing mechanisms have been identified. A new mechanism was proposed by Garner and Greenwood as a potential contribution to IASCC that involves the radiation-induced release into solution of sulphur and other deleterious elements that are normally concentrated into MnS precipitates. The instability arises from the combined action of the transmutation of manganese to iron, cascade-induced mixing and the very strong action of the inverse Kirkendall effect. The latter mechanism acts as a pump to export manganese from the precipitate surface and to replace it primarily with iron, as well as smaller amounts of chromium, nickel and other lesser elements. Evidence previously presented by Chung and coworkers appears to show that MnS precipitates in typical 300 series stainless steels become progressively depleted in manganese and enriched with iron as irradiation proceeds in boiling water reactor neutron spectra. It is shown in this paper that transmutation alone is insufficient to produce the observed behavior.
Date: October 1, 1997
Creator: Garner, F. A.; Greenwood, L. R. & Chung, H. M.
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

Description: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.
Date: May 1, 1997
Creator: Chopra, O.K.; Chung, H.M. & Gruber, E.E.
Partner: UNT Libraries Government Documents Department