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

On a Thermal Analysis of a Second Stripper for Rare Isotope Accelerator.

This memo summarizes simple calculations and results of the thermal analysis on the second stripper to be used in the driver linac of Rare Isotope Accelerator (RIA). Both liquid (Sodium) and solid (Titanium and Vanadium) stripper concepts were considered. These calculations were intended to provide basic information to evaluate the feasibility of liquid (thick film) and solid (rotating wheel) second strippers. Nuclear physics calculations to estimate the volumetric heat generation in the stripper material were performed by 'LISE for Excel'. In the thermal calculations, the strippers were modeled as a thin 2D plate with uniform heat generation within the beam spot. Then, temperature distributions were computed by assuming that the heat spreads conductively in the plate in radial direction without radiative heat losses to surroundings

Development of stripper options for FRIB

The US Department of Energy Facility for Rare Isotope Beams (FRIB) at Michigan State University includes a heavy ion superconducting linac capable of accelerating all ions up to uranium with energies higher than 200 MeV/u and beam power up to 400 kW. To achieve these goals with present ion source performance it is necessary to accelerate simultaneously two charge states of uranium from the ion source in the first section of the linac. At an energy of approximately 16.5 MeV/u it is planned to strip the uranium beam to reduce the voltage needed in the rest of the linac to achieve the final energy. Up to five different charge states are planned to be accelerated simultaneously after the stripper. The design of the stripper is a challenging problem due to the high power deposited (approximately 0.7 kW) in the stripper media by the beam in a small spot. To assure success of the project we have established a research and development program that includes several options: carbon or diamond foils, liquid lithium films, gas strippers and plasma strippers. We present in this paper the status of the different options

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

Advanced Burner Test Reactor Preconceptual Design Report.

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible

Advances of the FRIB project

The Facility for Rare Isotope Beams (FRIB) Project has entered the phase of beam commissioning starting from the room-temperature front end and the superconducting linac segment of first 15 cryomodules. With the newly commissioned helium refrigeration system supplying 4.5K liquid helium to the quarter-wave resonators and solenoids, the FRIB accelerator team achieved the sectional key performance parameters as designed ahead of schedule accelerating heavy ion beams above 20MeV/u energy. Thus, FRIB accelerator becomes world's highest-energy heavy ion linear accelerator. We also validated machine protection and personnel protection systems that will be crucial to the next phase of commissioning. FRIB is on track towards a national user facility at the power frontier with a beam power two orders of magnitude higher than operating heavy-ion facilities. This paper summarizes the status of accelerator design, technology development, construction, commissioning as well as path to operations and upgrades

Factors influencing overall survival rates for patients with pineocytoma

Given its rarity, appropriate treatment for pineocytoma remains variable. As the literature primarily contains case reports or studies involving a small series of patients, prognostic factors following treatment of pineocytoma remain unclear. We therefore compiled a systematic review of the literature concerning post-treatment outcomes for pineocytoma to better determine factors associated with overall survival among patients with pineocytoma. We performed a comprehensive search of the published English language literature to identify studies containing outcome data for patients undergoing treatment for pineocytoma. Kaplan–Meier analysis was utilized to determine overall survival rates. Our systematic review identified 168 total patients reported in 64 articles. Among these patients, 21% underwent biopsy, 38% underwent subtotal resection, 42% underwent gross total resection, and 29% underwent radiation therapy, either as mono- or adjuvant therapy. The 1 and 5 year overall survival rates for patients receiving gross total resection versus subtotal resection plus radiotherapy were 91 versus 88%, and 84 versus 17%, respectively. When compared to subtotal resection alone, subtotal resection plus radiation therapy did not offer a significant improvement in overall survival. Gross total resection is the most appropriate treatment for pineocytoma. The potential benefit of conventional radiotherapy for the treatment of these lesions is unproven, and little evidence supports its use at present

Research proposal for development of an electron stripper using a thin liquid lithium film for rare isotope accelerator.

Hydrodynamic instability phenomena in a thin liquid lithium film, which has been proposed for the first stripper in the driver linac of Rare Isotope Accelerator (RIA), were discussed. Since it was considered that film instability could significantly impair the feasibility of the liquid lithium film stripper concept, potential issues and research tasks in the RIA project due to these instability phenomena were raised. In order to investigate these instability phenomena, a research proposal plan was developed. In the theoretical part of this research proposal, a use of the linear stability theory was suggested. In the experimental part, it was pointed out that the concept of Reynolds number and Weber number scaling may allow conducting a preliminary experiment using inert simulants, hence reducing technical difficulty, complexity, and cost of the experiments. After confirming the thin film formation in the preliminary experiment using simulants, demonstration experiments using liquid lithium were proposed

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