Corrosion performance of advanced structural materials in sodium.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium

Creep rupture testing of alloy 617 and A508/533 base metals and weldments.

The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 750-1000 C. Alloy 617 is a prime candidate for VHTR structural components such as reactor internals, piping, and heat exchangers in view of its resistance to oxidation and elevated temperature strength. However, lack of adequate data on the performance of the alloy in welded condition prompted to initiate a creep test program at Argonne National Laboratory. In addition, Testing has been initiated to evaluate the creep rupture properties of the pressure vessel steel A508/533 in air and in helium environments. The program, which began in December 2009, was certified for quality assurance NQA-1 requirements during January and February 2010. Specimens were designed and fabricated during March and the tests were initiated in April 2010. During the past year, several creep tests were conducted in air on Alloy 617 base metal and weldment specimens at temperatures of 750, 850, and 950 C. Idaho National Laboratory, using gas tungsten arc welding method with Alloy 617 weld wire, fabricated the weldment specimens. Eight tests were conducted on Alloy 617 base metal specimens and nine were on Alloy 617 weldments. The creep rupture times for the base alloy and weldment tests were up to {approx}3900 and {approx}4500 h, respectively. The results showed that the creep rupture lives of weld specimens are much longer than those for the base alloy, when tested under identical test conditions. The test results also showed that the creep strain at fracture is in the range of 7-18% for weldment samples and were much lower than those for the base alloy, under similar test conditions. In general, the weldment specimens showed more of a flat or constant creep rate region than the base metal specimens. The base alloy and the weldment exhibited tertiary creep after 50-60% of the rupture life, irrespective of test temperature in the range of 750-950 C. The results showed that the stress dependence of the creep rate followed a power law for both base alloy and weldments. The data also showed that the stress exponent for creep is the same and one can infer that the same mechanism is operative in both base metal and weldments in the temperature range of the current study. SEM fractography analysis indicated that both base metal and weldment showed combined fracture modes consisting of dimple rupture and intergranular cracking. Intergranular cracking was more evident in the weldment specimens, which is consistent with the observation of lower creep ductility in the weldment than in the base metal

Report on sodium compatibility of advanced structural materials.

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 550 C; the weight loss of ferritic-martensitic steels, G92 and G91 is more significant than that of austenitic stainless steel, HT-UPS after sodium exposures at 650 C. Sodium exposures up to 2700 h at 550 C had no significant influence on tensile properties, while sodium exposures up to 5064 h at 650 C dramatically lowered the tensile strengths of the four alloys. The ultimate tensile strength of H1 G92, H2 G92, and G91 ferritic-martensitic steels was reduced to as much as nearly half of its initial value after sodium exposures at 650 C. Though the uniform elongation was recovered to some extent, these three ferritic-martensitic steels showed considerable strain softening after sodium exposures. The yield stress of HT-UPS austenitic stainless steel increased, the ultimate tensile strength decreased, and the total elongation was reduced after sodium exposures at 650 C. The dynamic strain aging effect observed in the as-received HT-UPS specimens became less pronounced after sodium exposures at 650 C. Microstructural characterization of sodium-exposed specimens showed no appreciable surface deterioration or grain structure changes under an optical microscope, except for the H2 G92 steel, in which the martensite structure transformed to large grain ferrite after sodium exposures at 650 C. TEM observations of the sodium-exposed H2 G92 steel showed significant recrystallization after sodium exposure for 2700 h at 550 C, and transformation of martensite to ferrite and high density of precipitates in nearly dislocation-free matrix after sodium exposures at 650 C. Further microstructural analysis and evaluation of decarburization/carburization behavior is needed to understand the dramatic changes in the tensile strengths of advanced ferritic-martensitic and austenitic steels after sodium exposures at 650 C

Creep Rupture Testing of Alloy 617 and A508/533 Base Metals and Weldments.

The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 750-1000 C. Alloy 617 is a prime candidate for VHTR structural components such as reactor internals, piping, and heat exchangers in view of its resistance to oxidation and elevated temperature strength. However, lack of adequate data on the performance of the alloy in welded condition prompted to initiate a creep test program at Argonne National Laboratory. In addition, Testing has been initiated to evaluate the creep rupture properties of the pressure vessel steel A508/533 in air and in helium environments. The program, which began in December 2009, was certified for quality assurance NQA-1 requirements during January and February 2010. Specimens were designed and fabricated during March and the tests were initiated in April 2010. During the past year, several creep tests were conducted in air on Alloy 617 base metal and weldment specimens at temperatures of 750, 850, and 950 C. Idaho National Laboratory, using gas tungsten arc welding method with Alloy 617 weld wire, fabricated the weldment specimens. Eight tests were conducted on Alloy 617 base metal specimens and nine were on Alloy 617 weldments. The creep rupture times for the base alloy and weldment tests were up to {approx}3900 and {approx}4500 h, respectively. The results showed that the creep rupture lives of weld specimens are much longer than those for the base alloy, when tested under identical test conditions. The test results also showed that the creep strain at fracture is in the range of 7-18% for weldment samples and were much lower than those for the base alloy, under similar test conditions. In general, the weldment specimens showed more of a flat or constant creep rate region than the base metal specimens. The base alloy and the weldment exhibited tertiary creep after 50-60% of the rupture life, irrespective of test temperature in the range of 750-950 C. The results showed that the stress dependence of the creep rate followed a power law for both base alloy and weldments. The data also showed that the stress exponent for creep is the same and one can infer that the same mechanism is operative in both base metal and weldments in the temperature range of the current study. SEM fractography analysis indicated that both base metal and weldment showed combined fracture modes consisting of dimple rupture and intergranular cracking. Intergranular cracking was more evident in the weldment specimens, which is consistent with the observation of lower creep ductility in the weldment than in the base metal

Corrosion Performance of Advanced Structural Materials in Sodium.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium

Report on sodium compatibility of advanced structural materials.

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 550 C; the weight loss of ferritic-martensitic steels, G92 and G91 is more significant than that of austenitic stainless steel, HT-UPS after sodium exposures at 650 C. Sodium exposures up to 2700 h at 550 C had no significant influence on tensile properties, while sodium exposures up to 5064 h at 650 C dramatically lowered the tensile strengths of the four alloys. The ultimate tensile strength of H1 G92, H2 G92, and G91 ferritic-martensitic steels was reduced to as much as nearly half of its initial value after sodium exposures at 650 C. Though the uniform elongation was recovered to some extent, these three ferritic-martensitic steels showed considerable strain softening after sodium exposures. The yield stress of HT-UPS austenitic stainless steel increased, the ultimate tensile strength decreased, and the total elongation was reduced after sodium exposures at 650 C. The dynamic strain aging effect observed in the as-received HT-UPS specimens became less pronounced after sodium exposures at 650 C. Microstructural characterization of sodium-exposed specimens showed no appreciable surface deterioration or grain structure changes under an optical microscope, except for the H2 G92 steel, in which the martensite structure transformed to large grain ferrite after sodium exposures at 650 C. TEM observations of the sodium-exposed H2 G92 steel showed significant recrystallization after sodium exposure for 2700 h at 550 C, and transformation of martensite to ferrite and high density of precipitates in nearly dislocation-free matrix after sodium exposures at 650 C. Further microstructural analysis and evaluation of decarburization/carburization behavior is needed to understand the dramatic changes in the tensile strengths of advanced ferritic-martensitic and austenitic steels after sodium exposures at 650 C