WorldWideScience

Sample records for alloy-d-9

  1. Corrosion of alloy D9 in liquid sodium

    International Nuclear Information System (INIS)

    Alloy D9 (15Cr-15Ni-Mo-Ti-Si) is chosen as clad and wrapper material for 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. The successful operation of the reactor depends on the compatibility of the core structural materials such as clad and wrapper which are subject to high neutron irradiation and in contact with high temperature liquid sodium. The chemical compatibility of sodium with the core structural materials is generally good when the sodium is in the pure state. One of the major criteria for selection of clad and wrapper materials is corrosion in liquid sodium environment. The corrosion of alloy D9 in presence of sodium was studied at different temperatures (773, 798, 823 and 873 K) for various duration of time ranging from 1000 to 3000 h in a well characterized sodium loop. The observed corrosion data such as weight loss, depleted layer formation, changes in microstructure of the alloy D9 specimens are compared with the literature data available on sodium corrosion in static and dynamic conditions. The corrosion rates in dynamic and static sodium are comparable in the measured temperatures as the purity of sodium with respect to oxygen concentration is comparable. The corrosion rate is faster at higher temperature and for longer duration of exposure to liquid sodium. Loss of thickness of material at 873 K is < 5 μm per year. Exposure to sodium causes selective dissolution of alloying elements such as nickel and chromium. The metal losses, thickness of corroded layers and changes in chemical composition were only marginal at a maximum temperature of 873 K. Hence alloy D9 is a good choice for clad and wrapper tube material. (author)

  2. High Temperature Heat Capacity of Alloy D9 Using Drop Calorimetry Based Enthalpy Increment Measurements

    Science.gov (United States)

    Banerjee, Aritra; Raju, S.; Divakar, R.; Mohandas, E.

    2007-02-01

    Alloy D9 is a void-swelling resistant nuclear grade austenitic stainless steel (SS) based on AISI type 316-SS in which titanium constitutes an added predetermined alloying composition. In the present study, the high-temperature enthalpy values of alloy D9 with three different titanium-to-carbon mass percent ratios, namely Ti/C = 4, 6, and 8, have been measured using inverse drop calorimetry in the temperature range from 295 to 1323 K. It is found that within the level of experimental uncertainty, the enthalpy values are independent of the Ti-C mass ratio. The temperature dependence of the isobaric specific heat C P is obtained by a linear regression of the measured enthalpy data. The measured C P data for alloy D9 may be represented by the following best-fit expression: C_P(J \\cdot kg^{-1}\\cdot K^{-1})= 431 + 17.7 × 10^{-2}T + 8.72 × 10^{-5}/T^2. It is found that the measured enthalpy and specific heat values exhibit good agreement with reported data on 316 and other related austenitic stainless steels.

  3. Chemical compatibility of sodium exposed alloy D9 with B4C

    International Nuclear Information System (INIS)

    The control rods in India's first prototype fast breeder reactor (PFBR) would consist of high-density B4C (containing 67% 10B isotope) pellets enclosed in alloy D9 clad tubes. Enriched boron carbide required for control rod applications in PFBR is being produced indigenously by reacting elemental boron with graphite at high temperatures. The gap between the control rod material and the clad is filled with liquid sodium as the pins are of vented type. At the reactor operating temperatures of 773-873 K, the chemical interaction between B4C and D9 clad material surface modified by liquid sodium needs to be investigated and understood. Towards this, out-of-pile accelerated chemical compatibility experiments were carried out and the SEM/EDS, XRD and XPS results of the studies carried out at 973 K for 5000 h are presented in this paper

  4. Thermal property characterization of a titanium modified austenitic stainless steel (alloy D9)

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Aritra [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Raju, S. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)]. E-mail: sraju@igcar.ernet.in; Divakar, R. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Mohandas, E. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Panneerselvam, G. [Fuel Chemistry Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Antony, M.P. [Fuel Chemistry Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2005-12-01

    The temperature dependence of lattice parameter and enthalpy increment of alloy D9, a titanium modified nuclear grade austenitic stainless steel were studied using high temperature X-ray diffraction and inverse drop calorimetry techniques, respectively. A smooth variation of the lattice parameter of the austenite with temperature was found. The instantaneous and mean linear thermal expansion coefficients at 1350 K were estimated to be 2.12 x 10{sup -5} K{sup -1} and 1.72 x 10{sup -5} K{sup -1}, respectively. The measured enthalpy data were made use of in estimating heat capacity, entropy and Gibbs energy values. The estimated isobaric heat capacity C {sub p} at 298 K was found to be 406 J kg{sup -1} K{sup -1}. An integrated theoretical analysis of the thermal expansion and enthalpy data was performed to obtain approximate values of bulk modulus as a function of temperature.

  5. End plug welding of PFBR fuel tubes with a 2.5 kW CW CO2 laser

    International Nuclear Information System (INIS)

    The end plug weld of the fuel tube of 500 MWe Prototype Fast Breeder Reactor (PFBR) involves dissimilar welding of alloy D9 (a 15Cr-15Ni-2Mo stainless steel) fuel clad tube with end plug made of AISI 316 M stainless steel. Being a non-contact process, laser welding does not generate active wastes in the form of used electrodes and defects like tungsten inclusions are completely eliminated. Completely austenitic mode of solidification associated with alloy D9 (due to its low Creq/Nieq ratio ≅ 1) makes this alloy particularly susceptible to solidification cracking. The present paper presents correlation of the microstructure of laser welds with welding parameters. The study demonstrated that sound end welds of PFBR fuel tube with crack resistant microstructure can be obtained by optimizing laser welding parameters, proper positioning of focused laser beam and ramping of laser power during welding. (author)

  6. Analysis of Titanium in D9 alloy- an inter comparison study and application to radioactive sample

    International Nuclear Information System (INIS)

    Titanium modified austenitic stainless steel alloy (D9) is well known for its swelling resistance under demanding reactor ambience. Chemical characterization of such important alloy carries much larger role in the quality control aspects. This work compares various analytical procedures and techniques envisaged to determine titanium concentration in D9 samples and further extending the analysis to radioactive D9 wrapper and clad of MOX test fuel sub assembly. (author)

  7. Larson-Miller correlation for the effect of thermal ageing on the yield strength of a cold worked 15Cr-15Ni-Ti modified austenitic stainless steel

    International Nuclear Information System (INIS)

    For 20% cold worked 15Cr-15Ni-Ti modified austenitic stainless steel (Alloy D9), the Larson-Miller parameter can be used to describe the effects of prior thermal exposures to different time-temperature combinations on the 0.2% yield stress σ YS, ultimate strength and total elongation in subsequent tensile tests at 300, 723 and 923 K. A single master plot for all the tensile test temperatures was obtained by plotting the Larson-Miller parameter against the ratio S YS=(σ YS of thermally aged material)/(σ YS of un-aged material) at identical tensile testing temperature

  8. Effect of Ti/C ratio and prior cold work on the tensile properties of 15Cr-15Ni-2.2Mo-Ti modified austenitic stainless steel

    International Nuclear Information System (INIS)

    The effect of Ti/C ratio and prior cold work on the tensile properties of 15Cr-15Ni-2.2Mo-Ti modified austenitic stainless steel (conforming to ASTMA-771/UNS S38660 and commonly referred to as alloy D-9) were studied in the temperature range 300-1023K at a nominal strain rate of 3.2x10-4s-1. The Ti/C ratio had a complex influence on the strength while it systematically showed an inverse relation with reduction in area. The influence of prior cold work on strength was found to be retained up to 773K. The various manifestations of dynamic strain ageing (DSA) such as serrated flow, ductility minimum, peaks/plateaus in the variation of yield strength, ultimate tensile strength, flow stress and work hardening rate with temperature were observed in the temperature range 573-873 K. While at temperatures below 773 K the tensile properties were influenced by DSA, above 773 K they were influenced by recovery. The strength and ductility values of alloy D-9 were comparable with those of similar grades of austenitic stainless steel and were found to meet the design requirements of the Prototype Fast Breeder Reactor. (author). 32 refs., 11 figs

  9. Development of materials and manufacturing technologies for Indian fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

  10. Development of materials and manufacturing technologies for Indian fast reactor programme

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required for testing

  11. Development of materials and fabrication technologies for sodium cooled fast reactor

    International Nuclear Information System (INIS)

    A comprehensive programme on materials development and fabrication technologies is being pursued at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India for improving the economic viability of fast nuclear power plants through extension of fuel burn-up and decreasing doubling time. IFAC-1, a modified version of Alloy D9 has been developed through optimisation of titanium, phosphorous and silicon contents for better swelling and creep resistances, allowing enhanced fuel burn-up. Manufacturing technology for oxide dispersion strengthened 9Cr-ferritic steel clad tube has been established, improving the creep resistance of an inherently void swelling resistant material. Studies on improved 9Cr-1Mo steel varieties for clad and wrapper applications for future metallic fuel reactors with reduced doubling times are under way. Other alloys developed include improved versions of 316LN SS as a structural material and type IV cracking resistant grade 91 steel for steam generator applications. These alloys require special welding consumables and procedures, technology for which has been indigenously developed. An overview of the work carried out at IGCAR will be presented. (author)

  12. Biaxial creep deformation behavior of Fe–14Cr–15Ni–Ti modified austenitic stainless steel fuel cladding tube for sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • Significant amounts of creep strain is observed in the axial and hoop directions. • Hoop strain is much higher than the axial strain. • Steady state hoop rate is lower than steady state axial rate. • Steady state hoop rate is comparable with creep rate evaluated from uniaxial tests. • Alloy D9 exhibits anisotropy in creep deformation. - Abstract: Twenty percent cold worked Fe–14Cr–15Ni–Ti modified austenitic stainless steel is used as the cladding tube material for the fuel pins of the Prototype Fast Breeder Reactor in India. Biaxial creep properties of the tubes have been studied at 973 K by carrying out creep tests by internally pressurizing the tubes. Hoop and axial components of creep strain were measured and found to be significantly different. For a given gas pressure, steady state hoop rate was higher than the axial rate. Steady state hoop and axial creep rates followed Norton's power law with the same stress exponent n = 7. Steady state hoop rates determined from biaxial creep tests agreed with the steady state creep rates determined from uniaxial creep tests. For a thin walled closed tube under internal pressure, significant axial deformation along with hoop deformation is indicative of anisotropic deformation of the material

  13. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    International Nuclear Information System (INIS)

    Six test assemblies similar in design to the FFTF driver assembly but employing the advanced alloy D9 in place of Type 316 stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of this design. A single pinhole-type breach was incurred in one of the high exposure tests after a peak fuel burnup of 155 MWd/kgM and peak fast neutron fluence of 25 x 1022 n/cm2 (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared to analytical evaluations. A revised swelling correlation for D9 Alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast fluence lifetime of 21.0 x 1022 n/cm2, or a peak fuel burnup greater than 120 MWd/kgM. The extended lifetime capability of this design would reduce fuel supply requirements for the FFTF by a third relative to the reference driver design

  14. Biaxial creep deformation behavior of Fe–14Cr–15Ni–Ti modified austenitic stainless steel fuel cladding tube for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mathew, M.D., E-mail: mathew@igcar.gov.in [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Ravi, S.; Vijayanand, V.D.; Latha, S. [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Dasgupta, Arup [Physical Metallurgy Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Laha, K. [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2014-08-15

    Highlights: • Significant amounts of creep strain is observed in the axial and hoop directions. • Hoop strain is much higher than the axial strain. • Steady state hoop rate is lower than steady state axial rate. • Steady state hoop rate is comparable with creep rate evaluated from uniaxial tests. • Alloy D9 exhibits anisotropy in creep deformation. - Abstract: Twenty percent cold worked Fe–14Cr–15Ni–Ti modified austenitic stainless steel is used as the cladding tube material for the fuel pins of the Prototype Fast Breeder Reactor in India. Biaxial creep properties of the tubes have been studied at 973 K by carrying out creep tests by internally pressurizing the tubes. Hoop and axial components of creep strain were measured and found to be significantly different. For a given gas pressure, steady state hoop rate was higher than the axial rate. Steady state hoop and axial creep rates followed Norton's power law with the same stress exponent n = 7. Steady state hoop rates determined from biaxial creep tests agreed with the steady state creep rates determined from uniaxial creep tests. For a thin walled closed tube under internal pressure, significant axial deformation along with hoop deformation is indicative of anisotropic deformation of the material.

  15. Electron microscopy an indispensable tool for knowledge based design and development of nuclear materials

    International Nuclear Information System (INIS)

    Development of materials for core components such as clad and wrapper for the Indian sodium cooled fast reactors and the plasma facing components in the ITER program has been a continuous indigenous effort involving a close collaboration between the designer, materials researcher and industry. In recent times there has been an intensive effort to design and develop new radiation resistant and high temperature materials which include the advanced austenitic and ferritic steels. An elaborate TEM investigation of the 20% Cold Worked SS316 austenitic stainless steel wrapper exposed to different damage levels from the Fast Breeder Test Reactor at Kalpakkam provided an in depth understanding on the mechanism of evolution of radiation induced phase changes and voids. The identification of η and G phases with unique microchemistry at 40 and 83 dpa respectively and the consequent depletion of beneficial elements Ni and Si from the matrix, resulted in precipitate associated voids and a high degree of volumetric swelling. This knowledge provided the impetus to develop alloy D9 and its variants with higher Ni, Ti and optimum amounts of Si and P where precipitation of fine stable Ti carbides/carbonitrides and phosphides imparts superior strength while the matrix precipitate interfaces act as defect sinks to control void swelling

  16. Microstructure evaluation of fourth generation reactor materials using modeling

    International Nuclear Information System (INIS)

    Modified 9Cr-1Mo steel is a primary candidate material for reactor pressure vessel (RPV) of Very High Temperature Gas-Cooled Reactor (VHTR) for hydrogen production. RPV is a pressure boundary component of great importance because its integrity is directly related to overall safety of nuclear power plant. Alloy D9 (15 Ni - 14Cr - 2 Mo + Si + Ti) with 20% cold work has been chosen for clad and wrapper, stainless steel (SS) 316LN for hot leg. (>673K) components, SS 304 LN for cold leg components (<673K) and modified 9Cr-1Mo steel for Steam Generators (SG) both Fast Breeder Reactor (FBR) and Prototype Fast Breeder Reactor (PFBR). Forgings are needed in SS316 LN and modified 9Cr-1Mo in various shapes and sizes. The characterization of bulk materials for thermal properties is very essential in monitoring the alloy processes. This is because, a knowledge of the physical and chemical properties of materials and the relations among them is of interest in the selection of materials when they are used in the design and manufacturing of devices particularly for atomic reactors. Studies on the relations between mechanical and thermal properties are of interest to the steel and metal industries as these would give useful information on the relation between hardness and thermal diffusivity (α) of steel. Recently Pena-Rodriíguez et al have determined the thermal diffusivity of low carbon steel at room temperature using the photo acoustic (PA) technique in a heat transmission configuration. Sivabharathy et al reported that photoacoustic (PA) measurements were carried out on modified 9Cr-1Mo steel to find out thermal diffusivity and thermal conductivity. Multiscale modeling approach is very useful for micro structure evaluation, thermal conductivity, thermal diffusivity and radiation effects. (author)

  17. Radiation-Induced Segregation and Phase Stability in Candidate Alloys for the Advanced Burner Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gary S. Was; Brian D. Wirth

    2011-05-29

    Major accomplishments of this project were the following: 1) Radiation induced depletion of Cr occurs in alloy D9, in agreement with that observed in austenitic alloys. 2) In F-M alloys, Cr enriches at PAG grain boundaries at low dose (<7 dpa) and at intermediate temperature (400°C) and the magnitude of the enrichment decreases with temperature. 3) Cr enrichment decreases with dose, remaining enriched in alloy T91 up to 10 dpa, but changing to depletion above 3 dpa in HT9 and HCM12A. 4) Cr has a higher diffusivity than Fe by a vacancy mechanism and the corresponding atomic flux of Cr is larger than Fe in the opposite direction to the vacancy flux. 5) Cr concentration at grain boundaries decreases as a result of vacancy transport during electron or proton irradiation, consistent with Inverse Kirkendall models. 6) Inclusion of other point defect sinks into the KLMC simulation of vacancy-mediated diffusion only influences the results in the low temperature, recombination dominated regime, but does not change the conclusion that Cr depletes as a result of vacancy transport to the sink. 7) Cr segregation behavior is independent of Frenkel pair versus cascade production, as simulated for electron versus proton irradiation conditions, for the temperatures investigated. 8) The amount of Cr depletion at a simulated planar boundary with vacancy-mediated diffusion reaches an apparent saturation value by about 1 dpa, with the precise saturation concentration dependent on the ratio of Cr to Fe diffusivity. 9) Cr diffuses faster than Fe by an interstitial transport mechanism, and the corresponding atomic flux of Cr is much larger than Fe in the same direction as the interstitial flux. 10) Observed experimental and computational results show that the radiation induced segregation behavior of Cr is consistent with an Inverse Kirkendall mechanism.

  18. Laser etching of austenitic stainless steels for micro-structural evaluation

    Science.gov (United States)

    Baghra, Chetan; Kumar, Aniruddha; Sathe, D. B.; Bhatt, R. B.; Behere, P. G.; Afzal, Mohd

    2015-06-01

    Etching is a key step in metallography to reveal microstructure of polished specimen under an optical microscope. A conventional technique for producing micro-structural contrast is chemical etching. As an alternate, laser etching is investigated since it does not involve use of corrosive reagents and it can be carried out without any physical contact with sample. Laser induced etching technique will be beneficial especially in nuclear industry where materials, being radioactive in nature, are handled inside a glove box. In this paper, experimental results of pulsed Nd-YAG laser based etching of few austenitic stainless steels such as SS 304, SS 316 LN and SS alloy D9 which are chosen as structural material for fabrication of various components of upcoming Prototype Fast Breeder Reactor (PFBR) at Kalpakkam India were reported. Laser etching was done by irradiating samples using nanosecond pulsed Nd-YAG laser beam which was transported into glass paneled glove box using optics. Experiments were carried out to understand effect of laser beam parameters such as wavelength, fluence, pulse repetition rate and number of exposures required for etching of austenitic stainless steel samples. Laser etching of PFBR fuel tube and plug welded joint was also carried to evaluate base metal grain size, depth of fusion at welded joint and heat affected zone in the base metal. Experimental results demonstrated that pulsed Nd-YAG laser etching is a fast and effortless technique which can be effectively employed for non-contact remote etching of austenitic stainless steels for micro-structural evaluation.

  19. Characterization of the mechanism of bi-layer oxide growth on austenitic stainless steels 316L and D9 in oxygen-controlled Lead-Bismuth Eutectic (LBE)

    Science.gov (United States)

    Koury, Daniel

    Lead Bismuth Eutectic (LBE) has been proposed for use in programs for accelerator-based and reactor-based transmutation of nuclear waste. LBE is a leading candidate material as a spallation target (in accelerator-based transmutation) and an option for the sub-critical blanket coolant. The corrosion by LBE of annealed and cold-rolled 316L stainless steels, and the modified austenitic stainless steel alloy D9, has been studied using Scanning Electron Microscopy (SEM), Electron Probe Micro Analysis (EPMA), and X-ray Photoelectron Spectroscopy (XPS). Exposed and unexposed samples have been compared and the differences studied. Small amounts of surface contamination are present on the samples and have been removed by ion-beam sputtering. The unexposed samples reveal typical stainless steel characteristics: a chromium oxide passivation surface layer and metallic iron and nickel. The exposed samples show protective iron oxide and chromium oxide growths on the surface. Oxygen takes many forms on the exposed samples, including oxides of iron and chromium, carbonates, and organic acids from subsequent handling after exposure to LBE. Different types of surface preparation have lead to considerably different modes of corrosion. The cold-rold samples were resistant to thick oxide growth, having only a thin (primary diffusant, as there are fewer fast-diffusion pathways and therefore an amount of chromium insufficient to maintain a chromium based oxide. Even the thick oxide, however, can prolong the life of a steel in LBE, provided proper oxygen control. The mechanisms responsible for the differences in the oxidation behaviors are discussed.

  20. Advanced materials for nuclear reactor systems: Alloys by design to overcome past limitations

    International Nuclear Information System (INIS)

    trade-offs must all be weighed carefully. Recently, a program to provide advanced structural materials for fast reactor applications was initiated within the United States. A thorough down-select process was conducted to weigh the requirements and benefits for all classes of structural materials. Four alloys were identified for further development. These include two ferritic-martensitic steels (NF616 and NF616 with special thermomechanical treatments) and two austenitic stainless steel alloys (high-temperature, ultrafine precipitation strengthened steel (HT-UPS) and NF709). NF616 is a 9Cr advanced ferritic/martensitic steel originally developed for super-critical boiler applications, while the HT-UPS alloys are 14Cr-16Ni austenitic stainless steels that were developed in the late 1980s by the U.S. Fusion Reactor Materials program for improved radiation-resistance. Common ferritic-martensitic and austenitic stainless steels such as HT9 and 316, respectively, are traditional and proven materials for sodium fast reactors. However, the selected alloys offer considerable improvements in strength and creep resistance over these more mature steels and yet maintain other critical properties at the same level. The superior performance and potential for improved reactor performance is illustrated for HT-UPS, D-9, and the traditional 316 SS in the figure below which shows the allowable operating regime in stress-temperature space. (Alloy D9 is an advanced austenitic steel that was developed during the United States National Cladding and Duct Development programin the 1970s and 1980s). The maximum stress limit at 50 to 550 deg. C (423 to 823 K) is defined as 1/3 of the ultimate tensile strength, which is a more conservative design limit than 2/3 of the yield stress for stainless steel. The stress limit at higher temperatures is defined as 2/3 of the creep rupture strength at 105 hours. The HT-UPS steel offers an additional 165 MPa over 316 SS at 500 deg. C. This increased strength