Science.gov

Sample records for irradiated uranium-molybdenum alloy

  1. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  2. Hot rolling of thick uranium molybdenum alloys

    SciTech Connect

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  3. Powder formation of {gamma} uranium-molybdenum alloys via hydration-dehydration

    SciTech Connect

    Vaz de Oliveira, Fabio Branco; Durazzo, Michelangelo; Fontenele Urano de Carvalho, Elita; Saliba-Silva, Adonis Marcelo; Gracher Riella, Humberto

    2008-07-15

    Gamma uranium-molybdenum alloys has been considered as fuel phase in plate type fuel elements for MTR reactors, mainly due to their acceptable performance under irradiation and metallurgical processing. To its use as a dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and one of the techniques considered at IPEN / CNEN - Brazil is HDH (hydration-dehydration). The alloys were produced by the induction melting technique, and samples were obtained from the alloys for the thermal treatments, under constant flow of hydrogen, for temperatures varying from 400 deg C to 600 deg C and times from 1 to 4 hours, followed by dehydration. A preliminary characterization of the powders was made and the curves of mass variation versus time were obtained and related to the powder characteristics. This paper describes the first results on the development of the technology to the powder formation of the (5 to 10) % weight molybdenum {gamma}-UMo alloys, and discusses some of its aspects, mainly those related to the {gamma} {yields} {alpha} equilibrium data. (author)

  4. Aqueous corrosion behavior of uranium-molybdenum alloys

    NASA Astrophysics Data System (ADS)

    Gardner, Levi D.

    Nuclear fuel characterization requires understanding of the various conditions to which materials are exposed in-reactor. One of these important conditions is corrosion, particularly that of fuel constituents. Therefore, corrosion behavior is of special interest and an essential part of nuclear materials characterization efforts. In support of the Office of Material Management and Minimization's Reactor Conversion Program, monolithic uranium-10 wt% molybdenum alloy (U-Mo) is being investigated as a low enriched uranium alternative to highly enriched uranium dispersion fuel currently used in domestic high performance research reactors. The aqueous corrosion behavior of U-Mo is being examined at Pacific Northwest National Laboratory (PNNL) as part of U-Mo fuel fabrication capability activity. No prior study adequately represents this behavior given the current state of alloy composition and thermomechanical processing methods, and research reactor water chemistry. Two main measurement techniques were employed to evaluate U-Mo corrosion behavior. Low-temperature corrosion rate values were determined by means of U-Mo immersion testing and subsequent mass-loss measurements. The electrochemical behavior of each processing condition was also qualitatively examined using the techniques of corrosion potential and anodic potentiodynamic polarization. Scanning electron microscopy (SEM) and optical metallography (OM) imagery and hardness measurements provided supplemental corrosion analysis in an effort to relate material corrosion behavior to processing. The processing effects investigated as part of this were those of homogenization heat treatment (employed to mitigate the effects of coring in castings) and sub-eutectoid heat treatment, meant to represent additional steps in fabrication (such as hot isostatic pressing) performed at similar temperatures. Immersion mass loss measurements and electrochemical results both showed very little appreciable difference between

  5. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET

    SciTech Connect

    Visser, A; Robert Pierce, R

    2007-06-07

    H-Canyon Engineering requested the Savannah River National Laboratory (SRNL) to perform two solvent extraction experiments using dissolved Super Kukla (SK) material. The SK material is an uranium (U)-molybdenum (Mo) alloy material of 90% U/10% Mo by weight with 20% 235U enrichment. The first series of solvent extraction tests involved a series of batch distribution coefficient measurements with 7.5 vol % tributylphosphate (TBP)/n-paraffin for extraction from 4-5 M nitric acid (HNO{sub 3}), using 4 M HNO{sub 3}-0.02 M ferrous sulfamate (Fe(SO3NH2)2) scrub, 0.01 M HNO3 strip steps with particular emphasis on the distribution of U and Mo in each step. The second set of solvent extraction tests determined whether the 2.5 wt % sodium carbonate (Na2CO3) solvent wash change frequency would need to be modified for the processing of the SK material. The batch distribution coefficient measurements were performed using dissolved SK material diluted to 20 g/L (U + Mo) in 4 M HNO{sub 3} and 5 M HNO{sub 3}. In these experiments, U had a distribution coefficient greater than 2.5 while at least 99% of the nickel (Ni) and greater than 99.9% of the Mo remained in the aqueous phase. After extraction, scrub, and strip steps, the aqueous U product from the strip contains nominally 7.48 {micro}g Mo/g U, significantly less than the maximum allowable limit of 800 {micro}g Mo/g U. Solvent washing experiments were performed to expose a 2.5 wt % Na2CO3 solvent wash solution to the equivalent of 37 solvent wash cycles. The low Mo batch distribution coefficient in this solvent extraction system yields only 0.001-0.005 g/L Mo extracted to the organic. During the solvent washing experiments, the Mo appears to wash from the organic.

  6. Effects of heat treatments on the thermal diffusivity of Uranium-Molybdenum alloy

    NASA Astrophysics Data System (ADS)

    Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Pedrosa, T. A.

    2016-07-01

    U-Mo alloys are the most investigated nuclear fuel material to be used in research reactors. The addition of molybdenum stabilizes the gamma phase of uranium and increases its melting point. A research program under development at Nuclear Technology Development Center (CDTN) aims the obtaining of uranium-molybdenum alloys to enable the high enriched uranium (HEU) to low enriched uranium (LEU) conversions. U-Mo ingots with 10% by weight were induction melted and heat treated at 300 °C for 72 h, 120 h and 240 h. Thermal diffusivity was determined by the laser flash method and thermal quadrupole method, from room temperature to 300 oC and 400oC. It was observed that the thermal diffusivity tends to increase with increasing temperature.

  7. Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

    NASA Astrophysics Data System (ADS)

    Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

    2009-03-01

    Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

  8. Uranium-Molybdenum Dissolution Flowsheet Studies

    SciTech Connect

    Pierce, R. A.

    2007-03-01

    The Super Kukla (SK) Prompt Burst Reactor operated at the Nevada Test Site from 1964 to 1978. The SK material is a uranium-molybdenum (U-Mo) alloy material of 90% U/10% Mo by weight at approximately 20% 235U enrichment. H-Canyon Engineering (HCE) requested that the Savannah River National Lab (SRNL) define a flowsheet for safely and efficiently dissolving the SK material. The objective is to dissolve the material in nitric acid (HNO3) in the H-Canyon dissolvers to a U concentration of 15-20 g/L (3-4 g/L 235U) without the formation of precipitates or the generation of a flammable gas mixture. Testing with SK material validated the applicability of dissolution and solubility data reported in the literature for various U and U-Mo metals. Based on the data, the SK material can be dissolved in boiling 3.0-6.0 M HNO3 to a U concentration of 15-20 g/L and a corresponding Mo concentration of 1.7-2.2 g/L. The optimum flowsheet will use 4.0-5.0 M HNO3 for the starting acid. Any nickel (Ni) cladding associated with the material will dissolve readily. After dissolution is complete, traditional solvent extraction flowsheets can be used to recover and purify the U. Dissolution rates for the SK material are consistent with those reported in the literature and are adequate for H-Canyon processing. When the SK material dissolved at 70-100 o C in 1-6 M HNO3, the reaction bubbled vigorously and released nitrogen oxide (NO) and nitrogen dioxide (NO2) gas. Gas generation tests in 1 M and 2 M HNO3 at 100 o C generated less than 0.1 volume percent hydrogen (H2) gas. It is known that higher HNO3 concentrations are less favorable for H2 production. All tests at 70-100 o C produced sufficient gas to mix the solutions without external agitation. At room temperature in 5 M HNO3, the U-Mo dissolved slowly and the U-laden solution sank to the bottom of the dissolution vessel because of its greater density. The effect of the density difference insures that the SK material cannot dissolve and

  9. CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

    SciTech Connect

    Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

  10. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to

  11. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  12. Development and validation of capabilities to measure thermal properties of layered monolithic U-Mo alloy plate-type fuel

    SciTech Connect

    Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-19

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of thermal conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify and validate the functionality of equipment methods installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, procedures to operate the equipment, and models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a zirconium diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  13. Effect of neutron irradiation on vanadium alloys

    SciTech Connect

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  14. Irradiation creep of dispersion strengthened copper alloy

    SciTech Connect

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  15. Irradiation creep of vanadium-base alloys

    SciTech Connect

    Tsai, H.; Billone, M.C.; Strain, R.V.; Smith, D.L.; Matsui, H.

    1998-03-01

    A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the United States. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200--300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 {times} 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

  16. Magnetic phase formation in irradiated austenitic alloys

    SciTech Connect

    Gussev, Maxim N; Busby, Jeremy T; Tan, Lizhen; Garner, Francis A.

    2014-01-01

    Austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with radiation-induced acceleration of the ferrite phase. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593K. The rate of development of magnetic phase appears to be sensitive to alloy composition. To the first order, the largest sensitivities to accelerate ferrite formation, as explored in this experiment, are associated with silicon, carbon and manganese and chromium. Si, C, and Mn are thought to influence diffusion rates of point defects while Cr plays a prominent role in defining the chromium equivalent and therefore the amount of ferrite at equilibrium. Pre-irradiation cold working was found to accelerate ferrite formation, but it can play many roles including an effect on diffusion, but on the basis of these results the dominant role or roles of cold-work cannot be identified. Based on the data available, ferrite formation is most probably associated with diffusion.

  17. Measurement of fission gas release from irradiated U-Mo monolithic fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  18. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

  19. High Temperature Irradiation Effects in Selected Generation IV Structural Alloys

    SciTech Connect

    Nanstad, Randy K; McClintock, David A; Hoelzer, David T; Tan, Lizhen; Allen, Todd R.

    2009-01-01

    In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 C and to irradiation doses in the range 1.2 to 1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 C; some hardening occurred at 660 C as well, but the 800H showed extremely low tensile elongations when tested at 700 C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.

  20. Slag remelt purification of irradiated vanadium alloys

    SciTech Connect

    Carmack, W.J.; Smolik, G.R.; McCarthy, K.A.; Gorman, P.K.

    1995-07-01

    This paper describes theoretical and scoping experimental efforts to investigate the decontamination potential of a slag remelting process for decontaminating irradiated vanadium alloys. Theoretical calculations, using a commercial thermochemical computer code HSC Chemistry, determined the potential slag compositions and slag-vanadium alloy ratios. The experiment determined the removal characteristics of four surrogate transmutation isotopes (Ca, Y - to simulate Sc, Mn, and Ar) from a V-5Ti-5Cr alloy with calcium fluoride slag. An electroslag remelt furnace was used in the experiment to melt and react the constituents. The process achieved about a 90 percent removal of calcium and over 99 percent removal of yttrium. Analyses indicate that about 40 percent of the manganese may have been removed. Argon analyses indicates that 99.3% of the argon was released from the vanadium alloy in the first melt increasing to 99.7% during the second melt. Powder metallurgy techniques were used to incorporate surrogate transmutation products in the vanadium. A powder mixture was prepared with the following composition: 90 wt % vanadium, 4.7 wt % titanium, 4.7 wt % chromium, 0.35 wt % manganese, 0.35 wt % CaO, and 0.35 wt % Y{sub 2}O{sub 3}. This mixture was packed into 2.54 cm diameter stainless steel tubes. Argon was introduced into the powder mixture by evacuating and backfilling the stainless steel containers to a pressure of 20 kPa (0.2 atm). The tubes were hot isostatically pressed at 207 MPa (2000 atm) and 1473 K to consolidate the metal. An electroslag remelt furnace (crucible dimensions: 5.1 cm diameter by 15.2 cm length) was used to process the vanadium electrodes. Chemical analyses were performed on samples extracted from the slags and ingots. Ingot analyses results are shown below. Values are shown in percent removal of the four targeted elements of the initial compositions.

  1. Nuclear fuel alloys or mixtures and method of making thereof

    DOEpatents

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  2. Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-12-31

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of {gamma}{double_prime} precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625.

  3. Alloy development for irradiation performance in fusion reactors

    NASA Astrophysics Data System (ADS)

    Harling, O. K.; Grant, N. J.

    1980-12-01

    The development of improved structural alloys for the fusion reactor first wall application is addressed. Several new alloys were produced by rapid solidification. Emphasis in alloy design and production was placed on producing austenitic Type 316SS with fine dispersions of TiC and Al2O3 particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations were initiated on samples fabricated from alloys produced. A dual beam, heavy ion, and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates, and total doses. Modeling of irradiation phenomena was continued with emphasis on understanding the effect of recoil resolution on relatively stable second phase particles. The microstructure of several ZrB2 doped stainless steels was characterized.

  4. Microstructures and Mechanical Properties of Irradiated Metals and Alloys

    SciTech Connect

    Zinkle, Steven J

    2008-01-01

    The effects of neutron irradiation on the microstructural evolution of metals and alloys are reviewed, with an emphasis on the roles of crystal structure, neutron dose and temperature. The corresponding effects of neutron irradiation on mechanical properties of metals and alloys are summarized, with particular attention on the phenomena of low temperature radiation hardening and embrittlement. The prospects of developing improved high-performance structural materials with high resistance to radiation-induced property degradation are briefly discussed.

  5. Swelling of several commercial alloys following high fluence neutron irradiation

    SciTech Connect

    Powell, R.W.; Peterson, D.T.; Zimmerschied, M.K.; Bates, J.F.

    1981-01-01

    Swelling values have been determined for a set of commercial alloys irradiated to a peak fluence of 17.8 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) over the temperature range of 400 to 650/sup 0/C. The alloys studied fall into three classes: the ferritic alloys AISI 430F, AISI 416, EM-12, H-11 and 2 1/4 Cr-1 Mo; the superalloys Inconel 718 and Inconel X-750; and the refractory alloys TZM and Nb-1 Zr. After irradiation to a peak fluence approaching goal exposures envisioned for advanced fusion reactor first walls, all of the alloys display swelling resistance far superior to cold worked AISI 316. Of the three alloy classes examined the swelling resistance of the ferritics is the least sensitive to composition.

  6. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  7. Modeling precipitate evolution in zirconium alloys during irradiation

    NASA Astrophysics Data System (ADS)

    Robson, J. D.

    2016-08-01

    The second phase precipitates (SPPs) in zirconium alloys are critical in controlling their performance. During service, SPPs are subject to both thermal and irradiation effects that influence volume fraction, number, and size. In this paper, a model has been developed to capture the combined effect of thermal and irradiation exposure on the Zr(Fe,Cr)2 precipitates in Zircaloy. The model includes irradiation induced precipitate destabilization integrated into a classical size class model for nucleation, growth and coarsening. The model has been applied to predict the effect of temperature and irradiation on SPP evolution. Increasing irradiation displacement rate is predicted to strongly enhance the loss of particles that arises from coarsening alone. The effect of temperature is complex due to competition between coarsening and irradiation damage. As temperature increases, coarsening is predicted to become increasingly important compared to irradiation induced dissolution and may increase resistance to irradiation induced dissolution by increasing particle size.

  8. Local phase transformation in alloys during charged-particle irradiation

    SciTech Connect

    Lam, N.Q.; Okamoto, P.R.

    1984-10-01

    Among the various mechanisms and processes by which energetic irradiation can alter the phase stability of alloys, radiation-induced segregation is one of the most important phenomena. Radiation-induced segregation in alloys occurs as a consequence of preferential coupling between persistent fluxes of excess defects and solute atoms, leading to local enrichment or depletion of alloying elements. Thus, this phenomenon tends to drive alloy systems away from thermodynamic equilibrium, on a local scale. During charged-particle irradiations, the spatial nonuniformity in the defect production gives rise to a combination of persistent defect fluxes, near the irradiated surface and in the peak-damage region. This defect-flux combination can modify the alloy composition in a complex fashion, i.e., it can destabilize pre-existing phases, causing spatially- and temporally-dependent precipitation of new metastable phases. The effects of radiation-induced segregation on local phase transformations in Ni-based alloys during proton bombardment and high-voltage electron-microscope irradiation at elevated temperatures are discussed.

  9. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1980

    SciTech Connect

    Not Available

    1981-04-01

    Progress is reported in eight sections: analysis and evaluation studies, test matrices and test methods development, Path A Alloy Development (austenitic stainless steels), Path C Alloy Development (Ti and V alloys), Path D Alloy Development (Fe alloys), Path E Alloy Development (ferritic steels), irradiation experiments and materials inventory, and materials compatibility and hydrogen permeation studies. (DLC)

  10. Radiation Damages in Aluminum Alloy SAV-1 under Neutron Irradiation

    NASA Astrophysics Data System (ADS)

    Salikhbaev, Umar; Akhmedzhanov, Farkhad; Alikulov, Sherali; Baytelesov, Sapar; Boltabaev, Azizbek

    2016-05-01

    The aim of this work was to study the effect of neutron irradiation on the kinetics of radiation damages in the SAV-1 alloy, which belongs to the group of aluminum alloys of the ternary system Al-Mg-Si. For fast-neutron irradiation by different doses up to fluence 1019 cm-2 the SAV-1 samples were placed in one of the vertical channels of the research WWR type reactor (Tashkent). The temperature dependence of the electrical resistance of the alloy samples was investigated in the range 290 - 490 K by the four-compensation method with an error about 0.1%. The experimental results were shown that at all the temperatures the dependence of the SAV-1 alloy resistivity on neutron fluence was nonlinear. With increasing neutron fluence the deviation from linearity and the growth rate of resistivity with temperature becomes more appreciable. The observed dependences are explained by means of martensitic transformations and the radiation damages in the studied alloy under neutron irradiation. The mechanisms of radiation modification of the SAV-1 alloy structure are discussed.

  11. Effects of self-irradiation in plutonium alloys

    DOE PAGES

    Chung, B. W.; Lema, K. E.; Allen, P. G.

    2015-09-16

    In this paper, we present updated results of self-irradiation effects on 238Pu-enriched 239Pu alloys measured by immersion density, dilatometry, and tensile tests. We obtained the self-irradiation equivalent time of nearly 200 years, nearly 100 years longer than in our previous papers. At this extended aging, we find the rate of decrease in density has slowed significantly, stabilizing around 15.73 g/cc, without signs of void swelling. The volume expansion measured at 35°C also shows apparent saturation at less than 0.25%. Quasi-static tensile measurement still show gradual increase in the strength of plutonium alloys with age.

  12. Self-irradiation of Pu, its alloys and compounds

    NASA Astrophysics Data System (ADS)

    Timofeeva, L. F.

    2000-07-01

    Self-irradiation of Pu, its alloys and compounds by products of known α-decomposition is a continuous complicated process, which includes numerous different phenomena. The accumulation of Pu decomposition products causes material structure and properties change. This problem is the subject of many works, most of them concerned with the behavior of Pu and its alloys at low (liquid He and N) temperatures. The survey is given of the results of our experiments connected with radiogenic helium behavior, crystal structure and properties of Pu metallic compounds and Pu oxide ceramics in a self-irradiation process at room temperature under isochronal heat treatments.

  13. Study of irradiation creep of vanadium alloys

    SciTech Connect

    Tsai, H.; Strain, R.V.; Smith, D.L.

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  14. Irradiation-assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-07-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water. New data confirms previous results that high irradiation levels reduce SCC resistance in Alloy X-750. Low boron heats show improved IASCC (irradiation-assisted stress corrosion cracking). Alloy 625 is resistant to IASCC. Microstructural, microchemical, and deformation studies were carried out. Irradiation of X-750 caused significant strengthening and ductility loss associated with formation of cavities and dislocation loops. High irradiation did not cause segregation in X-750. Irradiation of 625 resulted in formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to loops and precipitates was apparently offset in 625 by partial dissolution of {gamma} precipitates. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in X-750, and the absence of these two effects results in superior IASCC resistance in 625.

  15. Irradiation assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Mills, W.J.; Lebo, M.R.; Bajaj, R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.

    1994-06-01

    In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360{degree}C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K{sub 1} and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750.

  16. RERTR-13 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; M. A. Lillo; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

    2012-09-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  17. Effects of self-irradiation in plutonium alloys

    SciTech Connect

    Chung, B. W.; Lema, K. E.; Allen, P. G.

    2015-09-16

    In this paper, we present updated results of self-irradiation effects on 238Pu-enriched 239Pu alloys measured by immersion density, dilatometry, and tensile tests. We obtained the self-irradiation equivalent time of nearly 200 years, nearly 100 years longer than in our previous papers. At this extended aging, we find the rate of decrease in density has slowed significantly, stabilizing around 15.73 g/cc, without signs of void swelling. The volume expansion measured at 35°C also shows apparent saturation at less than 0.25%. Quasi-static tensile measurement still show gradual increase in the strength of plutonium alloys with age.

  18. A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

    NASA Astrophysics Data System (ADS)

    Van Kleeck, Melissa A.

    The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

  19. Irradiation testing of high density uranium alloy dispersion fuels

    SciTech Connect

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.

  20. Microstructural examination of irradiated vanadium alloys

    SciTech Connect

    Gelles, D.S.; Chung, H.M.

    1997-04-01

    Microstructural examination results are reported for a V-5Cr-5Ti unirradiated control specimens of heat BL-63 following annealing at 1050{degrees}C, and V-4Cr-4Ti heat BL-47 irradiated in three conditions from the DHCE experiment: at 425{degrees}C to 31 dpa and 0.39 appm He/dpa, at 600{degrees}C to 18 dpa and 0.54 appm He/dpa and at 600{degrees}C to 18 dpa and 4.17 appm He/dpa.

  1. Effect of cryogenic irradiation on NERVA structural alloys

    NASA Technical Reports Server (NTRS)

    Dixon, C. E.; Davidson, M. J.; Funk, C. W.

    1972-01-01

    Several alloys (Hastelloy X, AISI 347, A-286 bolts, Inconel 718, Al 7039-T63 and Ti-5Al-2.5Sn ELI) were irradiated in liquid nitrogen (140 R) to neutron fluences between 10 to the 17th power and 10 to the 19th power nvt (E greater than 1.0 Mev). After irradiation, tensile properties were obtained in liquid nitrogen without permitting any warmup except for some specimens which were annealed at 540 R. The usual trend of radiation damage typical for materials irradiated at and above room temperature was observed, such as an increase in strength and decrease in ductility. However, the damage at 140 R was greater because this temperature prevented the annealing of radiation-induced defects which occurs above 140 R.

  2. Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  3. The effect of alloying elements on the defect structural evolution in neutron irradiated Ni alloys

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Xu, Q.; Satoh, Y.; Ohkubo, H.; Kiritani, M.

    2000-12-01

    The effect of alloying elements, Si (-5.8%: the volume size factor in Ni), Ge (+14.76%) and Sn (+74.08%), on void swelling in neutron irradiated Ni at 573 K was studied by transmission electron microscope (TEM) observation and positron annihilation lifetime measurement. Neutron irradiation dose was changed widely from 0.001 to 0.4 dpa using two reactors, the Kyoto University reactor (KUR) and the Japan materials testing reactor (JMTR). Voids were observed in pure Ni by TEM even after very small irradiation dose of 0.001 dpa. With increasing dose, the density of voids did not change much while their size increased. The same tendency was observed in Ni-2at.%Ge. In Ni-2at.%Sn and Ni-2at.%Si, however, no voids were observed by TEM at a damage dose of 0.4 dpa. But positron lifetime measurement revealed the existence of microvoids at a medium dose of irradiation. When irradiation dose increased to 0.4 dpa in Ni-2at.%Si and 0.13 dpa in Ni-2at.%Sn, their existence was not detected. Suppression of microvoids in these alloys is discussed from the standpoint of solute point defect interactions.

  4. Carbon--silicon coating alloys for improved irradiation stability

    DOEpatents

    Bokros, J.C.

    1973-10-01

    For ceramic nuclear fuel particles, a fission product-retaining carbon-- silicon alloy coating is described that exhibits low shrinkage after exposure to fast neutron fluences of 1.4 to 4.8 x 10/sup 21/ n/cm/sup 2/ (E = 0.18 MeV) at irradiation temperatures from 950 to 1250 deg C. Isotropic pyrolytic carbon containing from 18 to 34 wt% silicon is co-deposited from a gaseous mixiure of propane, helium, and silane at a temperature of 1350 to 1450 deg C. (Official Gazette)

  5. TEM Examination of Advanced Alloys Irradiated in ATR

    SciTech Connect

    Jian Gan, PhD

    2007-09-01

    Successful development of materials is critical to the deployment of advanced nuclear power systems. Irradiation studies of candidate materials play a vital role for better understanding materials performance under various irradiation environments of advanced system designs. In many cases, new classes of materials have to be investigated to meet the requirements of these advanced systems. For applications in the temperature range of 500 800ºC which is relevant to the fast neutron spectrum burner reactors for the Global Nuclear Energy Partnership (GNEP) program, oxide dispersion strengthened (ODS) and ferritic martensitic steels (e.g., MA957 and others) are candidates for advanced cladding materials. In the low temperature regions of the core (<600ºC), alloy 800H, HCM12A (also called T 122) and HT 9 have been considered.

  6. AFIP-4 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

    2012-01-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  7. AFIP-4 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

    2011-09-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  8. Irradiation-induced patterning in dilute Cu-Fe alloys

    NASA Astrophysics Data System (ADS)

    Stumphy, B.; Chee, S. W.; Vo, N. Q.; Averback, R. S.; Bellon, P.; Ghafari, M.

    2014-10-01

    Compositional patterning in dilute Cu1-xFex (x ≈ 12%) induced by 1.8 MeV Kr+ irradiation was studied as a function of temperature using atom probe tomography. Irradiation near room temperature led to homogenization of the sample, whereas irradiation at 300 °C and above led to precipitation and macroscopic coarsening. Between these two temperatures the irradiated alloys formed steady state patterns of composition where precipitates grew to a fixed size. The size in this regime increased somewhat with temperature. It was also observed that the steady state concentrations of Fe in Cu matrix and Cu in the Fe precipitates both greatly exceeded their equilibrium solubilities, with the degree of supersaturation in each phase decreasing with increasing temperature. In the macroscopic coarsening regime, the Fe-rich precipitates showed indications of a “cherry-pit” structure, with Cu precipitates forming within the Fe precipitates. In the patterning regime, interfaces between Fe-rich precipitates and the Cu-rich matrix were irregular and diffuse.

  9. Complex nanoprecipitate structures induced by irradiation in immiscible alloy systems

    NASA Astrophysics Data System (ADS)

    Shu, Shipeng; Bellon, P.; Averback, R. S.

    2013-04-01

    We investigate the fundamentals of compositional patterning induced by energetic particle irradiation in model A-B substitutional binary alloys using kinetic Monte Carlo simulations. The study focuses on a type of nanostructure that was recently observed in dilute Cu-Fe and Cu-V alloys, where precipitates form within precipitates, a morphology that we term “cherry-pit” structures. The simulations show that the domain of stability of these cherry-pit structures depends on the thermodynamic and kinetic asymmetry between the A and B elements. In particular, both lower solubilities and diffusivities of A in B compared to those of B in A favor the stabilization of these cherry-pit structures for A-rich average compositions. The simulation results are rationalized by extending the analytic model introduced by Frost and Russell for irradiation-induced compositional patterning so as to include the possible formation of pits within precipitates. The simulations indicate also that the pits are dynamical structures that undergo nearly periodic cycles of nucleation, growth, and absorption by the matrix.

  10. Correlation between shear punch and tensile data for neutron-irradiated aluminum alloys

    SciTech Connect

    Hamilton, M.L.; Edwards, D.J.; Toloczko, M.B.

    1995-04-01

    This work was performed to determine whether shear punch and tensile data obtained on neutron irradiated aluminum alloys exhibited the same type of relationship as had been seen in other work and to assess the validity of extrapolating the results to proton-irradiated alloys. This work was also meant to be the first of a series of similar test matrices designed to determine whether the shear punch/tensile relationship varied or was the same for different alloy classes.

  11. Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

    SciTech Connect

    J. Gan; D. Keiser; D. Wachs; B. Miller; T. Allen; M. Kirk; J. Rest

    2009-11-01

    Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up to ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.

  12. Proton irradiation damage of an annealed Alloy 718 beam window

    SciTech Connect

    Bach, H. T.; Anderoglu, O.; Saleh, T. A.; Romero, T. J.; Kelsey, C. T.; Olivas, E. R.; Sencer, B. H.; Dickerson, P. O.; Connors, M. A.; John, K. D.; Maloy, S. A.

    2015-04-01

    Mechanical testing and microstructural analysis was performed on an Alloy 718 window that was in use at the Los Alamos Neutron Science Center (LANSCE) Isotope Production Facility (IPF) for approximately 5 years. It was replaced as part of the IPF preventive maintenance program. The window was transported to the Wing 9 hot cells at the Chemical and Metallurgical Research (CMR) LANL facility, visually inspected and 3-mm diameter samples were trepanned from the window for mechanical testing and microstructural analysis. Shear punch testing and optical metallography was performed at the CMR hot cells. The 1-mm diameter shear punch disks were cut into smaller samples to further reduce radiation exposure dose rate using Focus Ion Beam (FIB) and microstructure changes were analyzed using a Transmission Electron Microscopy (TEM). Irradiation doses were determined to be ~0.2–0.7 dpa (edge) to 11.3 dpa (peak of beam intensity) using autoradiography and MCNPX calculations. The corresponding irradiation temperatures were calculated to be ~34–120 °C with short excursion to be ~47–220 °C using ANSYS. Mechanical properties and microstructure analysis results with respect to calculated dpa and temperatures show that significant work hardening occurs but useful ductility still remains. The hardening in the lowest dose region (~0.2–0.7 dpa) was the highest and attributed to the formation of γ" precipitates and irradiation defect clusters/bubbles whereas the hardening in the highest dose region (~11.3 dpa) was lower and attributed mainly to irradiation defect clusters and some thermal annealing.

  13. Proton irradiation damage of an annealed Alloy 718 beam window

    DOE PAGES

    Bach, H. T.; Anderoglu, O.; Saleh, T. A.; Romero, T. J.; Kelsey, C. T.; Olivas, E. R.; Sencer, B. H.; Dickerson, P. O.; Connors, M. A.; John, K. D.; et al

    2015-04-01

    Mechanical testing and microstructural analysis was performed on an Alloy 718 window that was in use at the Los Alamos Neutron Science Center (LANSCE) Isotope Production Facility (IPF) for approximately 5 years. It was replaced as part of the IPF preventive maintenance program. The window was transported to the Wing 9 hot cells at the Chemical and Metallurgical Research (CMR) LANL facility, visually inspected and 3-mm diameter samples were trepanned from the window for mechanical testing and microstructural analysis. Shear punch testing and optical metallography was performed at the CMR hot cells. The 1-mm diameter shear punch disks were cutmore » into smaller samples to further reduce radiation exposure dose rate using Focus Ion Beam (FIB) and microstructure changes were analyzed using a Transmission Electron Microscopy (TEM). Irradiation doses were determined to be ~0.2–0.7 dpa (edge) to 11.3 dpa (peak of beam intensity) using autoradiography and MCNPX calculations. The corresponding irradiation temperatures were calculated to be ~34–120 °C with short excursion to be ~47–220 °C using ANSYS. Mechanical properties and microstructure analysis results with respect to calculated dpa and temperatures show that significant work hardening occurs but useful ductility still remains. The hardening in the lowest dose region (~0.2–0.7 dpa) was the highest and attributed to the formation of γ" precipitates and irradiation defect clusters/bubbles whereas the hardening in the highest dose region (~11.3 dpa) was lower and attributed mainly to irradiation defect clusters and some thermal annealing.« less

  14. Proton irradiation damage of an annealed Alloy 718 beam window

    NASA Astrophysics Data System (ADS)

    Bach, H. T.; Anderoglu, O.; Saleh, T. A.; Romero, T. J.; Kelsey, C. T.; Olivas, E. R.; Sencer, B. H.; Dickerson, P. O.; Connors, M. A.; John, K. D.; Maloy, S. A.

    2015-04-01

    Mechanical testing and microstructural analysis was performed on an Alloy 718 window that was in use at the Los Alamos Neutron Science Center (LANSCE) Isotope Production Facility (IPF) for approximately 5 years. It was replaced as part of the IPF preventive maintenance program. The window was transported to the Wing 9 hot cells at the Chemical and Metallurgical Research (CMR) LANL facility, visually inspected and 3-mm diameter samples were trepanned from the window for mechanical testing and microstructural analysis. Shear punch testing and optical metallography was performed at the CMR hot cells. The 1-mm diameter shear punch disks were cut into smaller samples to further reduce radiation exposure dose rate using Focus Ion Beam (FIB) and microstructure changes were analyzed using a Transmission Electron Microscopy (TEM). Irradiation doses were determined to be ∼0.2-0.7 dpa (edge) to 11.3 dpa (peak of beam intensity) using autoradiography and MCNPX calculations. The corresponding irradiation temperatures were calculated to be ∼34-120 °C with short excursion to be ∼47-220 °C using ANSYS. Mechanical properties and microstructure analysis results with respect to calculated dpa and temperatures show that significant work hardening occurs but useful ductility still remains. The hardening in the lowest dose region (∼0.2-0.7 dpa) was the highest and attributed to the formation of γ″ precipitates and irradiation defect clusters/bubbles whereas the hardening in the highest dose region (∼11.3 dpa) was lower and attributed mainly to irradiation defect clusters and some thermal annealing.

  15. Relationship of microstructure and tensile properties for neutron-irradiated vanadium alloys

    SciTech Connect

    Loomis, B.A.; Smith, D.L.

    1990-01-01

    The microstructures in V-15Cr-5Ti, V-10Cr-5RTi, V-3Ti-1Si, V-15Ti-7.5Cr, and V-20Ti alloys were examined by transmission electron microscopy after neutron irradiation at 600{degree}C to 21--84 atom displacements per atom in the Materials Open Test Assembly of the Fast Flux Test Facility. The microstructures in these irradiated alloys were analyzed to determine the radiation-produced dislocation density, precipitate number density and size, and void number density and size. The results of these analyses were used to compute increases in yield stress and swelling of the irradiated alloys. The computed increase in yield stress was compared with the increase in yield stress determined from tensile tests on these irradiated alloys. This comparison made it possible to evaluate the influence of alloy composition on the evolution of radiation-damaged microstructures and the resulting tensile properties. 11 refs.

  16. Effect of irradiation on the stress corrosion cracking behavior of Alloy X-750 and Alloy 625

    SciTech Connect

    Mills, W.J.; Lebo, M.R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.; Luther, R.F.; Sykes, G.B.

    1993-10-01

    In-reactor testing of bolt-loaded precracked and as-notched compact tension specimens was performed in 360{degrees}C water to determine effect of irradiation on SCC of Condition HTH and Condition BH Alloy X-750 and age-hardened Alloy 625. Variables were stress intensity factor (K{sub I}) level, fluence, grade of HTH material, prestraining and material chemistry. Effects of irradiation on high temperature SCC and the rapid cracking that occurs during cooldown below 150{degrees}C were characterized. Significant degradation in the in-reactor SCC resistance of HTH material was observed at initial K{sub I} levels above 30 MPa{radical}m and fluences greater than 10{sup 19} n/cm{sup 2} (E > 1 MeV). A small degradation in SCC resistance of HTH material was observed at low fluences (<10{sup 16} n/cm{sup 2}). As-notched specimens displayed less degradation in SCC resistance than precracked specimens. Prestraining greatly improved in-flux and out-of-flux SCC resistance of HTH material, as little or no SCC was observed in precracked specimens prestrained 20 to 30%, whereas extensive cracking was observed in nonprestrained specimens. Condition HTH heats with low boron (10 ppM or less) had improved in-reactor SCC resistance compared to heats with high and intermediate boron (>20 ppM). Age-hardened Alloy 625 exhibited superior in-reactor SCC behavior compared to HTH material as no crack extension occurred in any of the precracked Alloy 625 specimens tested at initial K{sub I} levels up to 80 MPa{radical}m.

  17. Void and precipitate structure in ion- and electron-irradiated ferritic alloys

    NASA Astrophysics Data System (ADS)

    Ohnuki, Soumei; Takahashi, Heishichiro; Takeyama, Taro

    1984-05-01

    Void formation and precipitation were investigated in Fe10Cr and Fe13Cr base alloys by 200 keV C + ion and 1 MeV electron irradiation. The ferritic alloys exhibited significant resistance to void swelling. In FeCr and FeCr-Si alloys, ion-irradiation produced the precipitates of M 23C 6 type. In the FeCrTi alloy, Ti-rich precipitates were formed with high number density on {100} plane. During electron-irradiation Fe-10Cr alloy, complex dislocation loops were produced with high number density, of which Burgers vector was mostly <100>. EDX analysis showed slightly enrichment of chromium on dislocation loops. These results suggested that the stability of <100> type dislocation structure at high dose is an important factor on good swelling resistance in the ferritic alloys, moreover, titanium addition will intensify the stability of the doslocations through the fine precipitation on dislocations.

  18. Microstructure evolution in austenitic Fe-Cr-Ni alloys irradiated with rotons: comparison with neutron-irradiated microstructures

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.

    2001-08-01

    Irradiation-induced microstructures of high purity and commercial purity austenitic stainless steels were investigated using proton-irradiation. For high purity alloys, Fe-20Cr-9Ni (HP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons between 300°C and 600°C at a dose rate of 7×10 -6 dpa/ s to doses up to 3.0 dpa. The commercial purity alloys, CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 5.0 dpa. The dose, temperature and composition dependence of the number density and size of dislocation loops and voids were characterized. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier-hardening (DBH) model. The dose and temperature dependence of proton-irradiated microstructure (loops, voids) and the irradiation hardening are consistent with the neutron-data trend. Results indicate that proton-irradiation can accurately reproduce the microstructure of austenitic alloys irradiated in LWR cores.

  19. Effect of irradiation on the tensile properties of niobium-base alloys

    SciTech Connect

    Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

    1986-11-01

    The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10SW neutrons/mS (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions.

  20. Effects of neutron irradiation on deformation behavior of nickel-base fastener alloys

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Kammenzind, B.F.; Burke, M.G.

    1999-07-01

    This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x10{sup 20} n/cm{sup 2} at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranular failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.

  1. Charpy impact test results for low activation ferritic alloys irradiated to 30 dpa

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1996-04-01

    Miniature specimens of six low activation ferritic alloys have been impact field tested following irradiation at 370{degrees}C to 30 dpa. Comparison of the results with those of control specimens and specimens irradiated to 10 dpa indicates that degradation in the impact behavior appears to have saturated by {approx}10 dpa in at least four of these alloys. The 7.5Cr-2W alloy referred to as GA3X appears most promising for further consideration as a candidate structural material in fusion reactor applications, although the 9Cr-1V alloy may also warrant further investigation.

  2. Kr ion irradiation study of the depleted-uranium alloys

    NASA Astrophysics Data System (ADS)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  3. RERTR-12 Insertion 2 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

    2012-09-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  4. Disassembly of irradiated lithium-bonded capsules containing vanadium alloy specimens

    SciTech Connect

    Tsai, H.; Strain, R.V.

    1996-04-01

    Capsules containing vanadium alloy specimens from irradiation experiments in FFTF and EBR-II are being processed to remove the lithium bond and retrieve the specimens for testing. The work has progressed smoothly.

  5. Evaluation of hardening behaviors in ion-irradiated Fe-9Cr and Fe-20Cr alloys by nanoindentation technique

    NASA Astrophysics Data System (ADS)

    Li, Shilei; Wang, Yanli; Dai, Xianyuan; Liu, Fang; Li, Jinyu; Wang, Xitao

    2016-09-01

    The ion irradiation hardening behaviors of Fe-9 wt% Cr and Fe-20 wt% Cr model alloys were investigated by nanoindentation technique. The specimens were irradiated with 3 MeV Fe11+ ions at room temperature up to 1 and 5 dpa for Fe-9Cr alloy and 1 and 2.5 for Fe-20Cr alloy. The ratio of average hardness in the same depth of irradiated and unirradiated (Hirr. av/Hunirr. av) was used to determine the critical indentation depth hcrit to eliminate the softer substrate effect. The Nix-Gao model was used to explain the indentation size effect. Irradiation hardening is clearly observed in both Fe-9Cr alloy and Fe-20Cr alloy after ion irradiation. The differences of ISE and irradiation hardening behaviors between Fe-9Cr and Fe-20Cr alloys are considered to be due to their different microstructures and microstructural evolution under ion irradiation.

  6. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    SciTech Connect

    Ashdown, B.G.

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

  7. Damage accumulation in ion-irradiated Ni-based concentrated solid-solution alloys

    DOE PAGES

    Ullah, Mohammad W.; Aidhy, Dilpuneet S.; Zhang, Yanwen; Weber, William J.

    2016-01-01

    We investigate Irradiation-induced damage accumulation in Ni0.8Fe0.2 and Ni0.8Cr0.2 alloys by using molecular dynamics simulations to assess possible enhanced radiation-resistance in these face-centered cubic (fcc), single-phase, concentrated solid-solution alloys, as compared with pure fcc Ni.

  8. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

    DOE PAGES

    Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-07-14

    The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition.more » Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.« less

  9. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

    SciTech Connect

    Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-07-14

    The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

  10. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

    NASA Astrophysics Data System (ADS)

    Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; Yamamoto, Yukinori; Snead, Lance L.

    2015-10-01

    The Fe-Cr-Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe-Cr-Al alloys has not been fully established. In this study, a series of Fe-Cr-Al alloys with 10-18 wt % Cr and 2.9-4.9 wt % Al were neutron irradiated at 382 °C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2<111> and a<100> were detected and quantified. Results indicate precipitation of Cr-rich α‧ is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure-property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α‧ precipitates at sufficiently high chromium contents after irradiation.

  11. Effects of alloying elements on the formation of < c >-component loops in Zr alloy Excel under heavy ion irradiation.

    SciTech Connect

    Idrees, Yasir; Francis, Elisabeth M.; Yao, Zhongwen; Korinek, Andreas; Kirk, Marquis A.; Sattari, Mohammad; Preuss, Michael; Daymond, M. R.

    2015-05-14

    We report here the microstructural changes occurring in the zirconium alloy Excel (Zr-3.5 wt% Sn-0.8Nb-0.8Mo-0.2Fe) during heavy ion irradiation. In situ irradiation experiments were conducted at reactor operating temperatures on two Zr Excel alloy microstructures with different states of alloying elements, with the states achieved by different solution heat treatments. In the first case, the alloying elements were mostly concentrated in the beta (beta) phase, whereas, in the second case, large Zr-3(Mo,Nb,Fe)(4) secondary phase precipitates (SPPs) were grown in the alpha (alpha) phase by long term aging. The heavy ion induced damage and resultant compositional changes were examined using transmission electron microscopy (TEM) in combination with scanning transmission electron microscope (STEM)-energy dispersive x-ray spectroscopy (EDS) mapping. Significant differences were seen in microstructural evolution between the two different microstructures that were irradiated under similar conditions. Nucleation and growth of < c >-component loops and their dependence on the alloying elements are a major focus of the current investigation. It was observed that the < c >-component loops nucleate readily at 100, 300, and 400 degrees C after a threshold incubation dose (TID), which varies with irradiation temperature and the state of alloying elements. It was found that the TID for the formation of < c >-component loops increases with decrease in irradiation temperature. Alloying elements that are present in the form of SPPs increase the TID compared to when they are in the beta phase solid solution. Dose and temperature dependence of loop size and density are presented. Radiation induced redistribution and clustering of alloying elements (Sn, Mo, and Fe) have been observed and related to the formation of < c >-component loops. It has been shown that at the higher temperature tests, irradiation induced dissolution of precipitates occurs whereas irradiation induced

  12. Mechanisms of Neutron Irradiation Hardening in Impurity-Doped Ferritic Alloys

    NASA Astrophysics Data System (ADS)

    Nishiyama, Y.; Liu, X. Y.; Kameda, J.

    2008-05-01

    Mechanisms of neutron irradiation hardening in phosphorus (P)-doped, sulfur (S)-doped, and copper (Cu)-doped ferritic alloys have been studied by applying a rate theory to the temperature dependence of the yield strength. Hardening behavior induced by neutron irradiation at various temperatures (473 to 711 K) is characterized in terms of the variations in athermal stress and activation energy for plasticity controlled by precipitation or solid solution, and kink-pair formation with the content and type of impurities. In P-doped alloys, neutron irradiation below 563 K brings about a remarkable increase in the athermal stress and activation energy, due to the dispersion of fine (˜1.7-nm) P-rich precipitates that is more extensive than that for the Cu-rich precipitates reported in irradiated steel. During neutron irradiation above 668 K, precipitation hardening occurs to some extent in Cu-doped and S-doped alloys, compared to small or negligible hardening in the P-doped alloys. In alloys with a low to moderate content of various dissolved impurities subjected to high-temperature irradiation, the formation of kink pairs becomes considerably difficult. Differing dynamic interactions of dissolved and precipitated impurities, i.e., P and Cu, with the nucleation and growth of dislocations are discussed, giving rise to irradiation hardening.

  13. Conceptual Design Parameters for HFIR LEU U-Mo Fuel Conversion Experimental Irradiations

    SciTech Connect

    Renfro, David G; Cook, David Howard; Chandler, David; Ilas, Germina; Jain, Prashant K

    2013-03-01

    The High Flux Isotope Reactor (HFIR) is a versatile research reactor that is operated at the Oak Ridge National Laboratory (ORNL). The HFIR core is loaded with high-enriched uranium (HEU) and operates at a power level of 85 MW. The primary scientific missions of the HFIR include cold and thermal neutron scattering, materials irradiation, and isotope production. An engineering design study of the conversion of the HFIR from HEU to low-enriched uranium (LEU) fuel is ongoing at the Oak Ridge National Laboratory. The LEU fuel considered is based on a uranium-molybdenum alloy that is 10 percent by weight molybdenum (U-10Mo) with a 235U enrichment of 19.75 wt %. The LEU core design discussed in this report is based on the design documented in ORNL/TM-2010/318. Much of the data reported in Sections 1 and 2 of this document was derived from or taken directly out of ORNL/TM-2010/318. The purpose of this report is to document the design parameters for and the anticipated normal operating conditions of the conceptual HFIR LEU fuel to aid in developing requirements for HFIR irradiation experiments.

  14. Microstructure evolution in proton-irradiated austenitic Fe-Cr-Ni alloys under LWR core conditions

    NASA Astrophysics Data System (ADS)

    Gan, Jian

    1999-11-01

    Irradiation-induced microstructure of austenitic stainless steel was investigated using proton irradiation. High-purity alloys of Fe-20Cr-9Ni (UHP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons at a dose rate of 7 × 10-6 dpa/s between 300°C and 600°C. The irradiation produced a microstructure consisting of dislocation loops and voids. The dose and temperature dependence of the number density and size of dislocation loops and voids were investigated. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier hardening model. The dose and temperature dependence of microstructure and hardness change for proton irradiation follows the same trend as that for neutron irradiation at comparable irradiation conditions. Commercial purity alloys of CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 3.0 dpa. The irradiated microstructure consists of dislocation loops. No voids were detected at doses up to 3.0 dpa. Loop size distributions are in close agreement with that in the same alloys neutron-irradiated in a LWR core. The loop density also agrees with neutron irradiation data. The yield strength as a function of dose in proton irradiated commercial purity alloys is consistent with the neutron- data trend. A fast-reactor microstructure model was adapted for light water reactor (LWR) irradiation conditions (275°C, 7 × 10 -8 dpa/s) and then applied to proton irradiation under conditions (360°C, 7 × 10-6 dpa/s) relevant to LWRs. The original model was modified by including in-cascade interstitial clustering and the loss of interstitial clusters to sinks by cluster diffusion. It was demonstrated that loop nucleation for both LWR irradiation condition and proton irradiation are driven by in-cascade interstitial clustering. One important result from this modeling work is that the difference in displacement cascade between

  15. Effect of fission neutron irradiation on the tensile and electrical properties of copper and copper alloys

    SciTech Connect

    Fabritsiev, S.A.; Zinkle, S.J.; Rowcliffe, A.F.

    1995-04-01

    The objective of this study is to evaluate the properties of several copper alloys following fission reactor irradiation at ITER-relevant temperatures of 80 to 200{degrees}C. This study provides some of the data needed for the ITER research and development Task T213. These low temperature irradiations caused significant radiation hardening and a dramatic decrease in the work hardening ability of copper and copper alloys. The uniform elongation was higher at 200{degree}C compared to 100{degree}C, but still remained below 1% for most of the copper alloys.

  16. Deuterium ion irradiation induced precipitation in Fe-Cr alloy: Characterization and effects on irradiation behavior

    NASA Astrophysics Data System (ADS)

    Liu, P. P.; Yu, R.; Zhu, Y. M.; Zhao, M. Z.; Bai, J. W.; Wan, F. R.; Zhan, Q.

    2015-04-01

    A new phase was found to precipitate in a Fe-Cr model alloy after 58 keV deuterium ion irradiation at 773 K. The nanoscale radiation-induced precipitate was studied systematically using high resolution transmission electron microscopy (HRTEM), image simulation and in-situ ultrahigh voltage transmission electron microscopy (HVEM). B2 structure is proposed for the new Cr-rich phase, which adopts a cube-on-cube orientation relationship with regard to the Fe matrix. Geometric phase analysis (GPA) was employed to measure the strain fields around the precipitate and this was used to explain its characteristic 1-dimensional elongation along the <1 0 0> Fe direction. The precipitate was stable under subsequent electron irradiation at different temperatures. We suggest that the precipitate with a high interface-to-volume ratio enhances the radiation resistance of the material. The reason for this is the presence of a large number of interfaces between the precipitate and the matrix, which may greatly reduce the concentration of point defects around the dislocation loops. This leads to a significant decrease in the growth rate.

  17. Ion irradiation induced disappearance of dislocations in a nickel-based alloy

    NASA Astrophysics Data System (ADS)

    Chen, H. C.; Li, D. H.; Lui, R. D.; Huang, H. F.; Li, J. J.; Lei, G. H.; Huang, Q.; Bao, L. M.; Yan, L.; Zhou, X. T.; Zhu, Z. Y.

    2016-06-01

    Under Xe ion irradiation, the microstructural evolution of a nickel based alloy, Hastelloy N (US N10003), was studied. The intrinsic dislocations are decorated with irradiation induced interstitial loops and/or clusters. Moreover, the intrinsic dislocations density reduces as the irradiation damage increases. The disappearance of the intrinsic dislocations is ascribed to the dislocations climb to the free surface by the absorption of interstitials under the ion irradiation. Moreover, the in situ annealing experiment reveals that the small interstitial loops and/or clusters induced by the ion irradiation are stable below 600 °C.

  18. Effect of neutron irradiation on fracture toughness behaviour of copper alloys

    NASA Astrophysics Data System (ADS)

    Tähtinen, S.; Pyykkönen, M.; Karjalainen-Roikonen, P.; Singh, B. N.; Toft, P.

    1998-10-01

    One of the most important factors in deciding about the applicability of materials in the structural components of ITER, is the effect of neutron irradiation on the fracture toughness behaviour of these materials. In the present work, the fracture toughness properties of two candidate materials for the first wall and divertor components of ITER, namely precipitation hardened CuCrZr and dispersion hardened CuAl25 alloys, have been studied in the unirradiated and irradiated conditions. In parallel, tensile properties of these alloys have been also investigated in the unirradiated and irradiated conditions.

  19. Irradiation effects in oxide dispersion strengthened (ODS) Ni-base alloys for Gen. IV nuclear reactors

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Ukai, Shigeharu; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2015-10-01

    Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m-3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

  20. Magnetic patterning using ion irradiation for highly ordered CoPt alloys with perpendicular anisotropy

    SciTech Connect

    Abes, M.; Venuat, J.; Muller, D.; Carvalho, A.; Schmerber, G.; Beaurepaire, E.; Dinia, A.; Pierron-Bohnes, V.

    2004-12-15

    We used a combination of ion irradiation and e-beam lithography to magnetically pattern an ordered CoPt alloy with strong perpendicular magnetic anisotropy. Ion irradiation disorders the alloy and strongly reduces the magnetic anisotropy. Magnetic force microscopy showed a regular array of 1 {mu}m{sup 2} square dots with perpendicular anisotropy separated by 1 {mu}m large ranges with in-plane anisotropy. This is further confirmed by magnetic measurements, which showed that arrays protected by a 200 nm Pt layer present the same coercive field and the same perpendicular anisotropy as before irradiation. This is promising for applications in magnetic recording technologies.

  1. Evolution of precipitate in nickel-base alloy 718 irradiated with argon ions at elevated temperature

    NASA Astrophysics Data System (ADS)

    Jin, Shuoxue; Luo, Fengfeng; Ma, Shuli; Chen, Jihong; Li, Tiecheng; Tang, Rui; Guo, Liping

    2013-07-01

    Alloy 718 is a nickel-base superalloy whose strength derives from γ'(Ni3(Al,Ti)) and γ″(Ni3Nb) precipitates. The evolution of the precipitates in alloy 718 irradiated with argon ions at elevated temperature were examined via transmission electron microscopy. Selected-area electron diffraction indicated superlattice spots disappeared after argon ion irradiation, which showing that the ordered structure of the γ' and γ″ precipitates became disordered. The size of the precipitates became smaller with the irradiation dose increasing at 290 °C.

  2. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    NASA Astrophysics Data System (ADS)

    Jin, K.; Bei, H.; Zhang, Y.

    2016-04-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm-2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.

  3. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    DOE PAGES

    Jin, Ke; Zhang, Yanwen; Bei, Hongbin

    2016-01-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm–2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing themore » ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.« less

  4. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    SciTech Connect

    Jin, Ke; Zhang, Yanwen; Bei, Hongbin

    2016-01-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm–2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.

  5. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    SciTech Connect

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  6. Microstructural evolution in nickel alloy C-276 after Ar-ion irradiation at elevated temperature

    SciTech Connect

    Jin, Shuoxue; He, Xinfu; Li, Tiecheng; Ma, Shuli; Tang, Rui; Guo, Liping

    2012-10-15

    In present work, the irradiation damage in nickel-base alloy C-276 irradiated with Ar-ions was studied. Specimens of C-276 alloy were subjected to an irradiation of Ar-ions (with 120 keV) to dose levels of 6 and 10 dpa at 300 and 550 Degree-Sign C, respectively. The size distributions and densities of dislocation loops caused by irradiation were investigated with transmission electron microscopy. Irradiation hardening due to the formation of the loops was calculated using the dispersed barrier-hardening model, showing that irradiation hardening was greatest at 300 Degree-Sign C/6 dpa. The microstructure evolution induced by Ar-ion irradiation (0-10 dpa) in nickel-base alloy C-276 has been studied using a multi-scale modeling code Radieff constructed based on rate theory, and the size of dislocation loops simulated by Radieff was in good agreement with the experiment. - Highlights: Black-Right-Pointing-Pointer High density of dislocation loops appeared after Ar ions irradiation. Black-Right-Pointing-Pointer Irradiation hardening due to the formation of loops was calculated by the DBH model. Black-Right-Pointing-Pointer Size of loops simulated by Radieff was in good agreement with the experiment.

  7. Phase Stability under Irradiation of Precipitates and Solid Solutions in Model ALloys and in ODS Alloys Relevant for Gen IV

    SciTech Connect

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

    The overall objective of this program is to investigate the irradiation-altered phase stability of oxide precipitates in ODS steels and of model alloy solid solutions of associated systems. This information can be used to determine whether the favorable mechanical propertiies of these steels are maintained under irradiation, thus addressing one of the main materials research issues for this class of steels as identified by the GenIV working groups. The research program will also create fundamental understanding of the irradiation precipitation/dissolution problem by studying a "model" system in which the variables can be controlled and their effects understood individually.

  8. Subtask 12F1: Effect of neutron irradiation on swelling of vanadium-base alloys

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the density change, void distribution, and microstructural evolution of vanadium-base alloys. Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420-600{degrees}C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti{sub 5}Si{sub 3} promotes good resistance to swelling of the Ti-containing alloys, and it was concluded that Ti of >3 wt.% and 400-1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. 18 refs., 6 figs., 1 tab.

  9. Subtask 12F3: Effects of neutron irradiation on tensile properties of vanadium-base alloys

    SciTech Connect

    Loomis, B.A.; Chung, H.M.; Smith, D.L.

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the tensile properties of candidate vanadium-base alloys. Vanadium-base alloys of the V-Cr-Ti system are attractive candidates for use as structural materials in fusion reactors. The current focus of the U.S. program of research on these alloys is on the V-(4-6)Cr-(3-6)Ti-(0.05-0.1)Si (in wt.%) alloys. In this paper, we present experimental results on the effects of neutron irradiation on tensile properties of selected candidate alloys after irradiation at 400{degrees}C-600{degrees}C in lithium in fast fission reactors to displacement damages of up to {approx}120 displacement per atom (dpa). Effects of irradiation temperature and dose on yield and ultimate tensile strengths and uniform and total elongations are given for tensile test temperatures of 25{degrees}C, 420{degrees}C, 500{degrees}, and 600{degrees}C. Effects of neutron damage on tensile properties of the U.S. reference alloy V-4Cr-4Ti are examined in detail. 7 refs., 10 figs., 1 tab.

  10. Microstructural Development in Irradiated U-7Mo/6061 Al Alloy Matrix Dispersion Fuel

    SciTech Connect

    Dennis D. Keiser, Jr.; Adam B. Robinson; Jan-Fong Jue; Pavel G. Medvedev; Daniel M. Wachs; M. Ross Finlay

    2009-09-01

    A U-7Mo alloy/6061 Al alloy matrix dispersion fuel plate was irradiated in the Advanced Test Reactor and then destructively examined using optical metallography and scanning electron microscopy to characterize the developed microstructure. Results were compared to the microstructures of as-fabricated dispersion fuel to identify changes that occurred during irradiation. The interaction layers that formed on the surface of the fuel U-7Mo particles during fuel fabrication exhibited stable irradiation performance as a result of the ~0.88 wt% Si present in the fuel meat matrix. During irradiation, the interaction layers changed very little in thickness and composition. The overall irradiation performance of the fuel plate to moderate power and burnup was considered excellent.

  11. Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys

    NASA Astrophysics Data System (ADS)

    Briggs, Samuel A.; Barr, Christopher M.; Pakarinen, Janne; Mamivand, Mahmood; Hattar, Khalid; Morgan, Dane D.; Taheri, Mitra; Sridharan, Kumar

    2016-10-01

    Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni4+ ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy.

  12. Dose dependence of mechanical properties in tantalum and tantalum alloys after low temperature irradiation

    SciTech Connect

    Byun, Thak Sang

    2008-01-01

    The dose dependence of mechanical properties was investigated for tantalum and tantalum alloys after low temperature irradiation. Miniature tensile specimens of three pure tantalum metals, ISIS Ta, Aesar Ta1, Aesar Ta2, and one tantalum alloy, Ta-1W, were irradiated by neutrons in the High Flux Isotope Reactor (HFIR) at ORNL to doses ranging from 0.00004 to 0.14 displacements per atom (dpa) in the temperature range 60 C 100 oC. Also, two tantalum-tungsten alloys, Ta-1W and Ta-10W, were irradiated by protons and spallation neutrons in the LANSCE facility at LANL to doses ranging from 0.7 to 7.5 dpa and from 0.7 to 25.2 dpa, respectively, in the temperature range 50 C 160 oC. Tensile tests were performed at room temperature and at 250oC at nominal strain rates of about 10-3 s-1. All neutron-irradiated materials underwent progressive irradiation hardening and loss of ductility with increasing dose. The ISIS Ta experienced embrittlement at 0.14 dpa, while the other metals retained significant necking ductility. Such a premature embrittlement in ISIS Ta is believed to be because of high initial oxygen concentrations picked up during a pre-irradiation anneal. The Ta-1W and Ta-10W specimens irradiated in spallation condition experienced prompt necking at yield since irradiation doses for those specimens were high ( 0.7 dpa). At the highest dose, 25.2 dpa, the Ta-10W alloy specimen broke with little necking strain. Among the test materials, the Ta-1W alloy displayed the best combination of strength and ductility. The plastic instability stress and true fracture stress were nearly independent of dose. Increasing test temperature decreased strength and delayed the onset of necking at yield.

  13. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

    NASA Astrophysics Data System (ADS)

    Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

    2016-08-01

    Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

  14. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    SciTech Connect

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  15. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    SciTech Connect

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  16. Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

    PubMed Central

    Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

    2016-01-01

    Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance. PMID:27562023

  17. Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

    NASA Astrophysics Data System (ADS)

    Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

    2016-08-01

    Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance.

  18. Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation.

    PubMed

    Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

    2016-01-01

    Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance. PMID:27562023

  19. Damage accumulation in ion-irradiated Ni-based concentrated solid-solution alloys

    SciTech Connect

    Ullah, Mohammad W.; Aidhy, Dilpuneet S.; Zhang, Yanwen; Weber, William J.

    2016-01-01

    We investigate Irradiation-induced damage accumulation in Ni0.8Fe0.2 and Ni0.8Cr0.2 alloys by using molecular dynamics simulations to assess possible enhanced radiation-resistance in these face-centered cubic (fcc), single-phase, concentrated solid-solution alloys, as compared with pure fcc Ni.

  20. Stress corrosion cracking behavior of irradiated model austenitic stainless steel alloys.

    SciTech Connect

    Chung, H. M.; Karlsen, T. M.; Ruther, W. E.; Shack, W. J.; Strain, R. V.

    1999-07-16

    Slow-strain-rate tensile tests (SSRTs) and posttest fractographic analyses by scanning electron microscopy were conducted on 16 austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} and {approx}0.9 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E >1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. Following irradiation to a fluence of {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2}, a high-purity laboratory heat of Type 316L SS (Si {approx} 0.024 wt%) exhibited the highest susceptibility to IGSCC. The other 15 alloys exhibited negligible susceptibility to IGSCC at this low fluence. The percentage of TGSCC on the fracture surfaces of SSRT specimens of the 16 alloys at {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E > 1 MeV) could be correlated well with N and Si concentrations; all alloys that contained <0.01 wt.% N and <1.0 wt. % Si were susceptible, whereas all alloys that contained >0.01 wt.% N or >1.0 wt.% Si were relatively resistant. High concentrations of Cr were beneficial. Alloys that contain <15.5 wt.% Cr exhibited greater percentages of TGSCC and IGSCC than those alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  1. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  2. Irradiation creep behavior of V-4Cr-4Ti alloys irradiated in a liquid sodium environment at the JOYO fast reactor

    NASA Astrophysics Data System (ADS)

    Fukumoto, Ken-ichi; Matsui, Hideki; Narui, Minoru; Yamazaki, Masanori

    2013-06-01

    Irradiation experiments on V-4Cr-4Ti alloys with sodium-enclosed irradiation capsules in the JOYO fast reactor were conducted using pressurized creep tubes (PCTs). The irradiation creep strain was significantly larger than the thermal creep strain below 686 °C, but there was no swelling of the neutron-irradiated V-4Cr-4Ti alloys. At temperatures below 500 °C, the irradiation creep was found to be proportional to the square root of the neutron dose and linear with the stress level. Above 500 °C, it was expected to be proportional to the stress level to a power greater than unity, because the irradiation creep mechanism could change from the stress-induced preferred absorption mechanism (SIPA) to the preferred absorption glide mechanism (PGA). By comparing annealed PCT specimens with cold-worked specimens, the cold-worked V-4Cr-4Ti alloys exhibited a larger irradiation creep strain compared with the annealed alloys. The irradiation creep compliance of the V-4Cr-4Ti alloys were ˜10 × 10-6 MPa-1 dpa-1 below 500 °C and 50-200 × 10-6 MPa-1 dpa-1 above 500 °C, a value greater than that of commercial V-4Cr-4Ti alloys, austenitic steels and ferritic steels.

  3. Nanostructure evolution under irradiation in FeMnNi alloys: A "grey alloy" object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Malerba, L.; Becquart, C. S.

    2015-07-01

    This work extends the object kinetic Monte Carlo model for neutron irradiation-induced nanostructure evolution in Fe-C binary alloys developed in [1], introducing the effects of substitutional solutes like Mn and Ni. The objective is to develop a model able to describe the nanostructural evolution of both vacancy and self-interstitial atom (SIA) defect cluster populations in Fe(C)MnNi neutron-irradiated model alloys at the operational temperature of light water reactors (∼300 °C), by simulating specific reference irradiation experiments. To do this, the effects of the substitutional solutes of interest are introduced, under simplifying assumptions, using a "grey alloy" scheme. Mn and Ni solute atoms are not explicitly introduced in the model, which therefore cannot describe their redistribution under irradiation, but their effect is introduced by modifying the parameters that govern the mobility of both SIA and vacancy clusters. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proved to be key to explain the experimentally observed disappearance of detectable defect clusters with increasing solute content. Solute concentration is explicitly taken into account in the model as a variable determining the slowing down of self-interstitial clusters; small vacancy clusters, on the other hand, are assumed to be significantly slowed down by the presence of solutes, while for clusters bigger than 10 vacancies their complete immobility is postulated. The model, which is fully based on physical considerations and only uses a few parameters for calibration, is found to be capable of reproducing the experimental trends in terms of density and size distribution of the irradiation-induced defect populations with dose, as compared to the reference experiment, thereby providing insight into the physical mechanisms that influence the nanostructural evolution undergone by this material during irradiation.

  4. Effect of neutron irradiation on the tensile properties and microstructure of several vanadium alloys

    SciTech Connect

    Braski, D.N.

    1986-01-01

    Specimens of V-15Cr-5Ti, VANSTAR-7, and V-3Ti-1Si were encapsulated in TZM tubes containing /sup 7/Li to prevent interstitial pickup and irradiated in FFTF (MOTA experiment) to a damage level of 40 dpa. The irradiation temperatures were 420, 520, and 600/sup 0/C. For a better simulation of fusion reactor conditions, helium was preimplanted in some specimens using a modified version of the ''tritium trick.'' The V-15Cr-5Ti alloy was most susceptible to irradiation hardening and helium embrittlement, followed by VANSTAR-7 and V-3Ti-1Si. VANSTAR-7 exhibited a relatively high maximum void swelling of approx.6% at 520/sup 0/C while V-15Cr-5Ti and V-3Ti-1Si had values of less than 0.3% at all three temperatures. The V-3Ti-1Si clearly outperformed the other two vanadium alloys in resisting the effects of neutron irradiation.

  5. Effects of compositional complexity on the ion-irradiation induced swelling and hardening in Ni-containing equiatomic alloys

    DOE PAGES

    Jin, K.; Lu, C.; Wang, L. M.; Qu, J.; Weber, W. J.; Zhang, Y.; Bei, H.

    2016-04-14

    The impact of compositional complexity on the ion-irradiation induced swelling and hardening is studied in Ni and six Ni-containing equiatomic alloys with face-centered cubic structure. The irradiation resistance at the temperature of 500 °C is improved by controlling the number and, especially, the type of alloying elements. Alloying with Fe and Mn has a stronger influence on swelling reduction than does alloying with Co and Cr. Lastly, the quinary alloy NiCoFeCrMn, with known excellent mechanical properties, has shown 40 times higher swelling tolerance than nickel.

  6. Effects of irradiation to 4 dpa at 390 C on the fracture toughness of vanadium alloys

    SciTech Connect

    Gruber, E.E.; Galvin, T.M.; Chopra, O.K.

    1998-09-01

    Fracture toughness J-R curve tests were conducted at room temperature on disk-shaped compact-tension DC(T) specimens of three vanadium alloys having a nominal composition of V-4Cr-4Ti. The alloys in the nonirradiated condition showed high fracture toughness; J{sub IC} could not be determined but is expected to be above 600 kJ/m{sup 2}. The alloys showed very poor fracture toughness after irradiation to 4 dpa at 390 C, e.g., J{sub IC} values of {approx}10 kJ/m{sup 2} or lower.

  7. Processing of Refractory Metal Alloys for JOYO Irradiations

    SciTech Connect

    RF Luther; ME Petrichek

    2006-02-21

    This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang.

  8. The effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 100 C

    SciTech Connect

    Edwards, D.J.; Singh, B.N.; Toft, P.; Eldrup, M.

    1998-03-01

    This report describes the final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. The post-irradiation tests at 100 C revealed the greatest loss of ductility occurred in the CuCrZr alloys, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which overall exhibited a factor of 3 higher uniform elongation after irradiation with almost double the strength. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The Al25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure. The results of this experiment confirm that the al25 possesses the most resistant microstructure to thermal and irradiation-induced changes, while the competing effects of ballistic dissolution and reprecipitation lead to important changes in the two precipitation strengthened alloys. This study and others have repeatedly shown that these materials can only be used if the very low uniform elongation (1% or less) can be accounted for in the design since pre-irradiation thermal processing cannot mitigate the irradiation embrittlement.

  9. Ion irradiation testing and characterization of FeCrAl candidate alloys

    SciTech Connect

    Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew; Wang, Yongqiang

    2014-10-29

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

  10. Hydrogen isotope transfer in austenitic steels and high-nickel alloy during in-core irradiation

    SciTech Connect

    Polosukhin, B.G.; Sulimov, E.M.; Zyrianov, A.P.; Kalinin, G.M.

    1995-10-01

    The transfer of protium and deuterium in austenitic chromium-nickel steels and in a high-nickel alloy was studied in a specially designed facility. The transfer parameters of protium and deuterium were found to change greatly during in-core irradiation, and the effects of irradiation increased as the temperature decreased. Thus, at temperature T<673K, the relative increase in the permeability of hydrogen isotopes under irradiation can be orders of magnitude higher in these steels. Other radiation effects were also observed, in addition to the changes from the initial values in the effects of protium and deuterium isotopic transfer. 4 refs., 3 figs., 2 tabs.

  11. Resistivity measurements of neutron-irradiated pure metals and Al-Zn alloys

    NASA Technical Reports Server (NTRS)

    Horak, J. A.

    1968-01-01

    Report presents resistivity measurements and their interpretation for neutron-irradiated pure metals and Al-Zn alloys. The influence of temperature, the role of point defects, and the aging behavior on resistivity are considered. The experimental procedures and results are discussed in detail.

  12. High post-irradiation ductility thermomechanical treatment for precipitation strengthened austenitic alloys

    DOEpatents

    Laidler, James J.; Borisch, Ronald R.; Korenko, Michael K.

    1982-01-01

    A method for improving the post-irradiation ductility is described which prises a solution heat treatment following which the materials are cold worked. They are included to demonstrate the beneficial effect of this treatment on the swelling resistance and the ductility of these austenitic precipitation hardenable alloys.

  13. 84 MeV C-ions irradiation effects on Zr-45Ti-5Al-3V alloy

    NASA Astrophysics Data System (ADS)

    Wang, Weipeng; Li, Zhengcao; Zhang, Zhengjun; Zhang, Chonghong

    2014-09-01

    Newly developed Zr-45Ti-5Al-3V alloy were irradiated by 84 MeV carbon ions with doses of 4 * 1015 ions/cm2 and 12 * 1015 ions/cm2, respectively. XRD, SEM, TEM, SAD and tensile tests were performed to study the microstructural evolution and mechanical properties modification upon high energy carbon ion irradiation. XRD patterns show no phase change while the diffraction peak position and intensity vary with irradiation doses. Tensile tests verify monotonic change of alloy strengths and elongations upon irradiation. Microstructure observations of the irradiated samples reveal the irradiation-induced precipitation of (Zr,Ti)3C2, which was believed contributing to the alloy hardening. Superlattice was discovered by the SAD patterns of original and irradiated samples and the high energy C-ions implantation was demonstrated to promote the disorder-order transition by introducing lattice defects.

  14. Irradiation-enhanced α' precipitation in model FeCrAl alloys

    DOE PAGES

    Edmondson, Philip D.; Briggs, Samuel A.; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar; Terrani, Kurt A.; Field, Kevin G.

    2016-02-17

    Model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) have been neutron irradiated at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich α' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. This is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning of the Al from themore » α' precipitates was also observed.« less

  15. Tensile properties of vanadium alloys irradiated at 390{degrees}C in EBR-II

    SciTech Connect

    Chung, H.M.; Tsai, H.C.; Nowicki, L.J.

    1997-08-01

    Vanadium alloys were irradiated in Li-bonded stainless steel capsules to {approx}390{degrees}C in the EBR-II X-530 experiment. This report presents results of postirradiation tests of tensile properties of two large-scale (100 and 500 kg) heats of V-4Cr-Ti and laboratory (15-30 kg) heats of boron-doped V-4Cr-4Ti, V-8Cr-6Ti, V-5Ti, and V-3Ti-1Si alloys. Tensile specimens, divided into two groups, were irradiated in two different capsules under nominally similar conditions. The 500-kg heat (No. 832665) and the 100-kg heat (VX-8) of V-4Cr-4Ti irradiated in one of the subcapsules exhibited complete loss of work-hardening capability, which was manifested by very low uniform plastic strain. In contrast, the 100-kg heat of V-4Cr-4Ti irradiated in another subcapsule exhibited good tensile properties (uniform plastic strain 2.8-4.0%). A laboratory heat of V-3Ti-1Si irradiated in the latter subcapsule also exhibited good tensile properties. These results indicate that work-hardening capability at low irradiation temperatures varies significantly from heat to heat and is influenced by nominally small differences in irradiation conditions.

  16. Mechanical properties and microstructural change of W–Y2O3 alloy under helium irradiation

    PubMed Central

    Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

    2015-01-01

    A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance. PMID:26227480

  17. U.S. Contribution 1994 Summary Report Task T12: Compatibility and irradiation testing of vanadium alloys

    SciTech Connect

    Smith, D.L.

    1995-03-01

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication.

  18. Ablation experiment and threshold calculation of titanium alloy irradiated by ultra-fast pulse laser

    SciTech Connect

    Zheng, Buxiang; Jiang, Gedong; Wang, Wenjun Wang, Kedian; Mei, Xuesong

    2014-03-15

    The interaction between an ultra-fast pulse laser and a material's surface has become a research hotspot in recent years. Micromachining of titanium alloy with an ultra-fast pulse laser is a very important research direction, and it has very important theoretical significance and application value in investigating the ablation threshold of titanium alloy irradiated by ultra-fast pulse lasers. Irradiated by a picosecond pulse laser with wavelengths of 1064 nm and 532 nm, the surface morphology and feature sizes, including ablation crater width (i.e. diameter), ablation depth, ablation area, ablation volume, single pulse ablation rate, and so forth, of the titanium alloy were studied, and their ablation distributions were obtained. The experimental results show that titanium alloy irradiated by a picosecond pulse infrared laser with a 1064 nm wavelength has better ablation morphology than that of the green picosecond pulse laser with a 532 nm wavelength. The feature sizes are approximately linearly dependent on the laser pulse energy density at low energy density and the monotonic increase in laser pulse energy density. With the increase in energy density, the ablation feature sizes are increased. The rate of increase in the feature sizes slows down gradually once the energy density reaches a certain value, and gradually saturated trends occur at a relatively high energy density. Based on the linear relation between the laser pulse energy density and the crater area of the titanium alloy surface, and the Gaussian distribution of the laser intensity on the cross section, the ablation threshold of titanium alloy irradiated by an ultra-fast pulse laser was calculated to be about 0.109 J/cm{sup 2}.

  19. Metastable phases in Zr-Excel alloy and their stability under heavy ion (Kr2+) irradiation

    NASA Astrophysics Data System (ADS)

    Yu, Hongbing; Zhang, Ken; Yao, Zhongwen; Kirk, Mark A.; Long, Fei; Daymond, Mark R.

    2016-02-01

    Zr-Excel alloy (Zr-3.5Sn-0.8Nb-0.8Mo, wt.%) has been proposed as a candidate material of pressure tubes in the CANDU-SCWR design. It is a dual-phase alloy containing primary hcp α-Zr and metastable bcc β-Zr. Metastable hexagonal ω-Zr phase could form in β-Zr as a result of aging during the processing of the tube. A synchrotron X-ray study was employed to study the lattice properties of the metastable phases in as-received Zr-Excel pressure tube material. In situ heavy ion (1 MeV Kr2+) irradiations were carried out at 200 °C and 450 °C to emulate the stability of the metastable phase under a reactor environment. Quantitative Chemi-STEM EDS analysis was conducted on both un-irradiated and irradiated samples to investigate alloying element redistribution induced by heavy ion irradiation. It was found that no decomposition of β-Zr was observed under irradiation at both 200 °C and 450 °C. However, ω-Zr particles experienced shape changes and shrinkage associated with enrichment of Fe at the β/ω interface during 200 °C irradiation but not at 450 °C. There is a noticeable increase in the level of Fe in the α matrix after irradiation at both 200 °C and 450 °C. The concentrations of Nb, Mo and Fe are increased in the ω phase but decreased in the β phase at 200 °C. The stability of metastable phases under heavy ion irradiation associated with elemental redistribution is discussed.

  20. Experimental study and numerical modelling of the irradiation damage recovery in zirconium alloys

    NASA Astrophysics Data System (ADS)

    Ribis, J.; Onimus, F.; Béchade, J.-L.; Doriot, S.; Barbu, A.; Cappelaere, C.; Lemaignan, C.

    2010-08-01

    Neutron irradiation damage in zirconium alloys used as fuel cladding tubes for Pressurized Water Reactors in the nuclear industry consists mainly in a high density of small prismatic dislocation loops. During post-irradiation heat treatment thermal annealing of loops occurs. This phenomenon has been investigated by transmission electron microscopy and microhardness tests. It has been shown that the loop density decreases while their mean size increases. Furthermore it was demonstrated that only vacancy loops remain present in the material after a long term annealing at high temperature. A mechanism based on vacancies diffusion has been proposed to explain the loop evolution during annealing. A cluster dynamic model, originally developed to compute the evolution of the microstructure under irradiation, has been adapted to the modelling of the annealing for zirconium alloys. This physically based model reproduces the loop size and density evolution during a large variety of heat treatments and also provides a better understanding of the mechanisms involved in the loop recovery.

  1. Results of the Irradiation of R6R018 in the Advanced Test Reactor

    SciTech Connect

    Adam B Robinson; Daniel Wachs; Pavel Medvedev; Curtis Clark; Gray Chang; Misti Lillo; Jan-Fong Jue; Glenn Moore; Jared Wight

    2010-04-01

    For over 30 years the Reduced Enrichment for Research and Test Reactors (RERTR) program has worked to provide the fuel technology and analytical support required to convert research and test reactors from nuclear fuels that utilize highly enriched uranium (HEU) to fuels based on low-enriched uranium (LEU) (defined as <20% U-235). This effort is driven by a desire to minimize international civilian commerce in weapons usable materials. The RERTR fuel development program has executed a wide array of fuel tests over the last decade that clearly established the viability of research reactor fuels based on uranium-molybdenum (U-Mo) alloys. Fuel testing has included a large number of dispersion type fuels capable of providing uranium densities up to approximately 8.5 g U/cc (~1.7 g U-235/cc at 20% enrichment). The dispersion fuel designs tested are very similar to existing research test reactor fuels in that the U-Mo particles simply replace the current fuel phase within the matrix. In 2003 it became evident that the first generation U-Mo-based dispersion fuel within an aluminum matrix exhibited significant fuel performance problems at high power and burn-up. These issues have been successfully addressed with a modest modification to the matrix material composition. Testing has shown that small additions of silicon (2–5 wt%) to the aluminum (Al) matrix stabilizes the fuel performance. The fuel plate R6R018 which was irradiated in the Advanced Test Reactor (ATR) as part of the RERTR-9B experiment was part of an investigation into the role of the silicon content in the matrix. This plate consisted of a U-7Mo fuel phase dispersed in an Al-3.5Si matrix clad in Al-6061. This report outlines the fabrication history, the as fabricated analysis performed prior to irradiation, the irradiation conditions, the post irradiation examination results, and an analysis of the plates behavior.

  2. Tensile properties of vanadium alloys irradiated at 200{degrees}C in the HFIR

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    1997-08-01

    Vanadium alloys were irradiated in a helium environment to {approx}10 dpa at {approx}200{degrees}C in the High Flux Isotope Reactor (HFIR). This report presents results of postirradiation tests of tensile properties of laboratory heats of V-(1-18)Ti, V-4Cr-4Ti, V-8Cr-6Ti, V-9Cr-5Ti, V-3Ti-1Si, and V-3Ti-0.1C alloys. Because of significant loss of work-hardening capability, all alloys except V-18Ti exhibited a very low uniform plastic strain <1%. For V-Ti. The mechanism of the loss of work-hardening capability in the other alloys is not understood.

  3. Irradiation damage behavior of low alloy steel wrought and weld materials

    SciTech Connect

    Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

    1993-10-01

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ``superclean`` composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature ({Delta}TT4{sub 41J} or {Delta}TT{sub 47J}) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest {Delta}TT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials.

  4. Investigation of Effects of Neutron Irradiation on Tantalum Alloys for Radioisotope Power System Applications

    SciTech Connect

    Barklay, Chadwick D.; Kramer, Daniel P.; Talnagi, Joseph

    2007-01-30

    Tantalum alloys have been used by the U.S. Department of Energy as structural alloys for space nuclear power systems such as Radioisotopic Thermoelectric Generators (RTG) since the 1960s. Tantalum alloys are attractive for high temperature structural applications due to their high melting point, excellent formability, good thermal conductivity, good ductility (even at low temperatures), corrosion resistance, and weldability. A number of tantalum alloys have been developed over the years to increase high-temperature strength (Ta-10%W) and to reduce creep strain (T-111). These tantalum alloys have demonstrated sufficient high-temperature toughness to survive the increasing high pressures of the RTG's operating environment resulting from the alpha decay of the 238-plutonium dioxide fuel. However, 238-plutonium is also a powerful neutron source. Therefore, the RTG operating environment produces large amounts of 3-helium and neutron displacement damage over the 30 year life of the RTG. The literature to date shows that there has been very little work focused on the mechanical properties of irradiated tantalum and tantalum alloys and none at the fluence levels associated with a RTG operating environment. The minimum, reactor related, work that has been reported shows that these alloys tend to follow trends seen in the behavior of other BCC alloys under irradiation. An understanding of these mechanisms is important for the confident extrapolation of mechanical-property trends to the higher doses and gas levels corresponding to actual service lifetimes. When comparing the radiation effects between samples of Ta-10%W and T-111 (Ta-8%W-2%Hf) subjected to identical neutron fluences and environmental conditions at temperatures <0.3Tm ({approx}700 deg. C), evidence suggests the possibility that T-111 will exhibit higher levels of internal damage accumulation and degradation of mechanical properties compared to Ta-10%W.

  5. Irradiation-induced phase transformations in zirconium alloys

    SciTech Connect

    Howe, L.M.; Phillips, D.; Motta, A.T.; Okamoto, P.R.

    1994-02-01

    {sup 40}Ar and {sup 209}Bi ion irradiations of Zr{sub 3}Fe were performed at 35--725 K using 15-1500 keV ions. Results are presented on role of deposited-energy density on nature of the damaged regions in individual cascades produced by ion bombardment of Zr{sub 3}Fe. Comparison is also made between irradiation-induced amorphization of Zr{sub 3}Fe during electron irradiation and under ion bombardments. Dependence of damage production on incident electron energy in Zr{sub 3}Fe was also determined. Preliminary results are also discussed for amorphization of ZrFe{sub 2}, Zr(Cr,Fe){sub 2} and ZrCr{sub 2} by electron irradiation. Results of a recent investigation on amorphization of Zr(Cr,Fe){sub 2} and Zr{sub 2}(Ni,Fe) precipitates in Zircaloy-4 are discussed in context of previous experimental results of neutron and electron irradiations and likely amorphization mechanisms are proposed.

  6. In situ HVEM studies of phase transformation in Zr alloys and compounds under irradiation

    SciTech Connect

    Motta, A.T.; Faldowski, J.A.; Howe, L.M.; Okamoto, P.R.

    1996-01-01

    The High Voltage Electron Microscope (HVEM)/Tandem facility at Argonne National Laboratory has been used to conduct detailed studies of the phase stability and microstructural evolution in zirconium alloys and compounds under ion and electron irradiation. Detailed kinetic studies of the crystalline-to-amorphous transformation of the intermetallic compounds Zr{sub 3}(Fe{sub 1-x}Ni{sub x}), Zr(Fe{sub 1-x},Cr{sub x}){sub 2}, Zr{sub 3}Fe, and Zr{sub 1.5} Nb{sub 1.5} Fe, both as second phase precipitates and in bulk form, have been performed using the in-situ capabilities of the Argonne facility, under a variety of irradiation conditions (temperature, dose rate). Results include a verification of a dose rate effect on amorphization and the influence of material variables (stoichiometry x, presence of stacking faults, crystal structure) on the critical temperature and on the critical dose for amorphization. Studies were also conducted of the microstructural evolution under irradiation of specially tailored binary and ternary model alloys. The stability of the {omega}-phase in Zr-20%Nb under electron and Ar ion irradiation was investigated as well as the {beta}-phase precipitation in Zr-2.5%Nb under Ar ion irradiation. The ensemble of these results is discussed in terms of theoretical models of amorphization and of irradiation-altered solubility.

  7. Tensile impact properties of vanadium-base alloys irradiated at <430{degree}C.

    SciTech Connect

    Chung, H. M.

    1998-05-18

    Tensile and impact properties were investigated at <430 C on V-Cr-Ti, V-Ti-Si, and V-Ti alloys after irradiation to {approx}2-46 dpa at 205-430 C in lithium or helium in the Fast Flux Test Facility (FFTF), High Flux Isotope Reactor (HFIR), Experimental Breeder Reactor II (EBR-II), and Advanced Test Reactor (ATR). A 500-kg heat of V-4Cr-4Ti exhibited high ductile-brittle transition temperature and minimal uniform elongation as a result of irradiation-induced loss of work-hardening capability. Work-hardening capabilities of 30- and 100-kg heats of V-4Cr-4Ti varied significantly with irradiation conditions, although the 30-kg heat exhibited excellent impact properties after irradiation at {approx}390-430 C. The origin of the significant variations in the work-hardening capability of V-4Cr-4Ti is not understood, although fabrication variables, annealing history, and contamination from the irradiation environment are believed to play important roles. A 15-kg heat of V-3Ti-1Si exhibited good work-hardening capability and excellent impact properties after irradiation at {approx}390-430 C. Helium atoms, either charged dynamically or produced via transmutation of boron in the alloys, promote work-hardening capability in V-4Cr-4Ti and V-3Ti-1Si.

  8. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  9. The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Munro, B.

    1997-04-01

    Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%) of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.

  10. Tensile properties of vanadium alloys irradiated at <430{degrees}C

    SciTech Connect

    Chung, H.M.; Smith, D.L.

    1997-08-01

    Recent attention to vanadium alloys has focused on significant susceptibility to loss of work-hardening capability in irradiation experiments at <430{degrees}C. An evaluation of this phenomenon was conducted on V-Ti, V-Cr-Ti, and V-Ti-Si alloys irradiated in several conventional and helium-charging irradiation experiments in the FFTF-MOTA, HFIR, and EBR-II. Work hardening capability and uniform tensile elongation appear to vary strongly from alloy and heat to heat. A strong heat-to-heat variation has been observed in V-4Cr-4Ti alloys tested, i.e., a 500-kg heat (No. 832665), a 100-kg heat (VX-8), and a 30-kg heat (BL-47). The significant differences in susceptibility to loss of work-hardening capability from one heat to another are estimated to correspond to a difference of {approx}100{degrees}C or more in minimum allowable operating temperature (e.g., 450 versus 350{degrees}C).

  11. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    NASA Astrophysics Data System (ADS)

    Idrees, Y.; Yao, Z.; Cui, J.; Shek, G. K.; Daymond, M. R.

    2016-11-01

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

  12. Helium effects on irradiation dmage in V alloys

    SciTech Connect

    Doraiswamy, N.; Alexander, D.

    1996-10-01

    Preliminary investigations were performed on V-4Cr-4Ti samples to observe the effects of He on the irradiation induced microstructural changes by subjecting 3 mm electropolished V-4Cr-4Ti TEM disks, with and without prior He implantation, to 200 keV He irradiation at room temperature and monitoring, in-situ, the microstructural evolution as a function of total dose with an intermediate voltage electron microscope directly connected to an ion implanter. A high density of black dot defects were formed at very low doses in both He pre-implanted and unimplanted samples.

  13. Voids in neutron-irradiated metals and alloys

    SciTech Connect

    Hendricks, R.W.

    1980-01-01

    Small-angle x-ray and neutron scattering are powerful analytical tools for investigating long-range fluctuations in electron (x-rays) or magnetic moment (neutrons) densities in materials. In recent years they have yielded valuable information about voids, void size distributions, and swelling in aluminum, aluminum alloys, copper, molybdenum, nickel, nickel-aluminum, niobium and niobium alloys, stainless steels, graphite and silicon carbide. In the case of aluminum, information concerning the shape of the voids and the ratio of specific surface energies was obtained. The technique of small-angle scattering and its application to the study of voids is reviewed in the paper. Emphasis is placed on the conditions which limit the applicability of the technique, on the interpretation of the data, and on a comparison of the results obtained with companion techniques such as transmission electron microscopy and bulk density. 8 figures, 41 references.

  14. Transmutations of elements under irradiation and its impact on alloys composition

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1994-09-01

    This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V-4Cr-4Ti), a reduced activation martensitic steel (Fe-9Cr-2WVTa), a high conductivity copper alloy (Cu-Cr-Zr), and the SiC compound. The fission irradiation facilities considered include the EBR-II (Experimental Breeder Reactor) fast reactor, and two high flux mixed spectrum reactors, HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The transmutation and dpa rates that occur in these test reactors are compared with the calculated transmutation and dpa rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions [(n,x{alpha}) and (n,xp)] are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from the ones predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time. As examples, the transmutation rates of W, Ta, V, Cu, among others, differ considerably when the irradiation is performed under a mixed spectrum reactor`s and fusion first wall`s spectrum. Fast reactors (EBR-II) provide the only possibility for obtaining high damage rates without producing significant compositional effects in vanadium alloys, ferritic steels and copper alloys.

  15. Microstructure analysis of magnesium alloy melted by laser irradiation

    NASA Astrophysics Data System (ADS)

    Liu, S. Y.; Hu, J. D.; Yang, Y.; Guo, Z. X.; Wang, H. Y.

    2005-12-01

    The effects of laser surface melting (LSM) on microstructure of magnesium alloy containing Al8.57%, Zn 0.68%, Mn0.15%, Ce0.52% were investigated. In the present work, a pulsed Nd:YAG laser was used to melt and rapidly solidify the surface of the magnesium alloy with the objective of changing microstructure and improving the corrosion resistance. The results indicate that laser-melted layer contains the finer dendrites and behaviors good resistance corrosion compared with the untreated layer. Furthermore, the absorption coefficient of the magnesium alloy has been estimated according to the numeral simulation of the thermal conditions. The formation process of fine microstructure in melted layers was investigated based on the experimental observation and the theoretical analysis. Some simulation results such as the re-solidification velocities are obtained. The phase constitutions of the melted layers determined by X-ray diffraction were β-Mg 17Al 12 and α-Mg as well as some phases unidentified.

  16. Impact properties of vanadium-base alloys irradiated at < 430 C

    SciTech Connect

    Chung, H.M.; Smith, D.L.

    1998-03-01

    Recent attention to vanadium-base alloys has focused on the effect of low-temperature (<430 C) neutron irradiation on the mechanical properties, especially the phenomena of loss of work-hardening capability under tensile loading and loss of dynamic toughness manifested by low impact energy and high ductile-brittle-transition temperature (DBTT). This paper summarizes results of an investigation of the low-temperature impact properties of V-5Ti, V-4Cr-4Ti, and V-3Ti-Si that were irradiated in several fission reactor experiments, i.e., FFTF-MOTA, EBR-II X-530, and ATR-A1. Irradiation performance of one production-scale and one laboratory heat of V-4C-4Ti and one laboratory heat of V-3Ti-Si was the focus of the investigation. Even among the same lass of alloy, strong heat-to-heat variation was observed in low-temperature impact properties. A laboratory heat of V-4Cr-4Ti and V-3Ti-1Si exhibited good impact properties whereas a 500-kg heat of V-4Cr-4Ti exhibited unacceptably high DBTT. The strong heat-to-heat variation in impact properties of V-4Cr-4Ti indicates that fabrication procedures and minor impurities play important roles in the low-temperature irradiation performance of the alloys.

  17. Irradiation Embritlement in Alloy HT-­9

    SciTech Connect

    Serrano De Caro, Magdalena

    2012-08-27

    HT-9 steel is a candidate structural and cladding material for high temperature lead-bismuth cooled fast reactors. In typical advanced fast reactor designs fuel elements will be irradiated for an extended period of time, reaching up to 5-7 years. Significant displacement damage accumulation in the steel is expected (> 200 dpa) when exposed to dpa-rates of 20-30 dpa{sub Fe}/y and high fast flux (E > 0.1 MeV) {approx}4 x 10{sup 15} n/cm{sup 2}s. Core temperatures could reach 400-560 C, with coolant temperatures at the inlet as low as 250 C, depending on the reactor design. Mechanical behavior in the presence of an intense fast flux and high dose is a concern. In particular, low temperature operation could be limited by irradiation embrittlement. Creep and corrosion effects in liquid metal coolants could set a limit to the upper operating temperature. In this report, we focus on the low temperature operating window limit and describe HT-9 embrittlement experimental findings reported in the literature that could provide supporting information to facilitate the consideration of a Code Case on irradiation effects for this class of steels in fast reactor environments. HT-9 has an extensive database available on irradiation performance, which makes it the best choice as a possible near-term candidate for clad, and ducts in future fast reactors. Still, as it is shown in this report, embrittlement data for very low irradiation temperatures (< 200 C) and very high radiation exposure (> 150 dpa) is scarce. Experimental findings indicate a saturation of DBTT shifts as a function of dose, which could allow for long lifetime cladding operation. However, a strong increase in DBTT shift with decreasing irradiation temperature could compromise operation at low service temperatures. Development of a deep understanding of the physics involved in the radiation damage mechanisms, together with multiscale computer simulation models of irradiation embrittlement will provide the basis to

  18. Alloy development for irradiation performance. Quarterly progress report for period ending September 30, 1980

    SciTech Connect

    Not Available

    1980-12-01

    This report is organized along topical lines in parallel to a Program Plan of the same title so that activities and accomplishments may be followed readily relative to that Program Plan. Thus, the work of a given laboratory may appear throughout the report. Chapters 1, 2, 8, and 9 review activities on analysis and evaluation, test methods development, status of irradiation experiments, and corrosion testing and hydrogen permeation studies, respectively. These activities relate to each of the alloy development paths. Chapters 3, 4, 5, 6, and 7 present the ongoing work on each alloy development path. The Table of Contents is annotated for the convenience of the reader.

  19. Alloy development for irradiation performance. Quarterly progress report for period ending March 31, 1980

    SciTech Connect

    Ashdown, B.G.

    1980-06-01

    This report is organized along topical lines in parallel to a Program Plan of the same title so that activities and accomplishments may be followed readily relative to that Program Plan. Thus, the work of a given laboratory may appear throughout the report. Chapters 1, 2, 8, and 9 review activities on analysis and evaluation, test methods development, status of irradiation experiments, and corrosion testing and hydrogen permeation studies, respectively. These activities relate to each of the alloy development paths. Chapters 3, 4, 5, 6, and 7 present the ongoing work on each alloy development path. The Table of Contents is annotated for the convenience of the reader.

  20. Alloy development for irradiation performance. Quarterly progress report for period ending June 30, 1980

    SciTech Connect

    Ashdown, B.G.

    1980-10-01

    This report is organized along topical lines in parallel to a Program Plan of the same title so that activities and accomplishments may be followed readily relative to that Program Plan. Thus, the work of a given laboratory may appear throughout the report. Chapters 1, 2, 8, and 9 review activities on analysis and evaluation, test methods development, status of irradiation experiments, and corrosion testing and hydrogen permeation studies, respectively. These activities relate to each of the alloy development paths. Chapters 3, 4, 5, 6, and 7 present the ongoing work on each alloy development path. The Table of Contents is annotated for the convenience of the reader.

  1. Response of nanostructured ferritic alloys to high-dose heavy ion irradiation

    SciTech Connect

    Parish, Chad M.; White, Ryan M.; LeBeau, James M.; Miller, Michael K.

    2014-02-01

    A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at -100°C and 750°C, as well as pre-irradiation reference material. Irradiation at -100°C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti-Y-O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750°C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr-W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

  2. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  3. Swelling and structure of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1994-08-01

    Combined effects of dynamically charged helium and neutron damage on density change, void distribution, and microstructural evolution of V-4Cr-4Ti alloy have been determined after irradiation to 18--31 dpa at 425--600 C in the Dynamic Helium Charging Experiment (DHCE), and the results were compared with those from a non-DHCE in which helium generation and negligible. For specimens irradiated to {approx}18-31 dpa at 500--600 with a helium generation rate of 0.4--4.2 appm He/dpa, only a few helium bubbles were observed at the interface of grain matrices and some of the Ti(O,N,C) precipitates, and no microvoids or helium bubbles were observed either in grain matrices or near grain boundaries. Under these conditions, dynamically produced helium atoms seem to be trapped in the grain matrix without significant bubble nucleation or growth, and in accordance with this, density changes from DHCE and non-DHCE (negligible helium generation) were similar for comparable fluence and irradiation temperature. Only for specimens irradiated to {approx}31 dpa at 425 C, when helium was generated at a rage of 0.4--0.8 appm helium/dpa, were diffuse helium bubbles observed in limited regions of grain matrices and near {approx}15% of the grain boundaries in densities significantly lower than those in the extensive coalescences of helium bubbles typical of other alloys irradiated in tritium-trick experiments. Density changes of specimens irradiated at 425 C in the DHCE were significantly higher than those from non-DHCE irradiation. Microstructural evolution in V-4Cr-4Ti was similar for DHCE and non-DHCE except for helium bubble number density and distribution. As in non-DHCE, the irradiation-induced precipitation of ultrafine Ti{sub 5}Si{sub 3} was observed for DHCE at >500 C but not at 425 C.

  4. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L.

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  5. Atom probe study of irradiation-enhanced α' precipitation in neutron-irradiated Fe–Cr model alloys

    SciTech Connect

    Chen, Wei -Ying; Miao, Yinbin; Wu, Yaqiao; Tomchik, Carolyn A.; Mo, Kun; Gan, Jian; Okuniewski, Maria A.; Maloy, Stuart A.; Stubbins, James F.

    2015-07-01

    Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on a' phase formation in Fe-Cr model alloys (10-16 at.%) irradiated at 300 and 450°C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α' precipitates with an average radius of 1.0-1.3 nm were observed. The precipitate density varied significantly from 1.1x10²³ to 2.7x10²⁴ 1/m³, depending on Cr concentrations and irradiation temperatures. The volume fraction of α' phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe-10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of a' precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.

  6. Effect of neutron-irradiation on the microstructure of a Fe-12at.%Cr alloy

    NASA Astrophysics Data System (ADS)

    Kuksenko, V.; Pareige, C.; Genevois, C.; Cuvilly, F.; Roussel, M.; Pareige, P.

    2011-08-01

    A nanoscale description of the microstructure in a Fe-12at%Cr model alloy of low purity which has been neutron irradiated at 300°C up to 0.6 dpa, has been performed owing to atom probe tomography (APT). APT investigations have shown that the impurities are also involved in the microstructural evolution under irradiation. Two different populations of clusters have been observed: Cr-enriched and NiSiPCr-enriched clusters. As expected with a process of enhanced precipitation, Cr-enriched clusters are homogeneously distributed inside grains and certainly not correlated to dislocation loops. The NiSiPCr-enriched clusters, which are probably radiation induced segregations, are independent of the Cr-enriched clusters and are also homogeneously distributed. A quantitative description of these objects is presented in this paper and results are compared to SANS data of the literature obtained for the same model alloy.

  7. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    NASA Astrophysics Data System (ADS)

    Sun, Yong; Huo, He-yong; Wu, Yang; Li, Jiangbo; Zhou, Wei; Guo, Hai-bing; Li, Hang; Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin

    2016-11-01

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 μm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  8. Phase stability and microstructures of high entropy alloys ion irradiated to high doses

    NASA Astrophysics Data System (ADS)

    Xia, Songqin; Gao, Michael C.; Yang, Tengfei; Liaw, Peter K.; Zhang, Yong

    2016-11-01

    The microstructures of AlxCoCrFeNi (x = 0.1, 0.75 and 1.5 in molar ratio) high entropy alloys (HEAs) irradiated at room temperature with 3 MeV Au ions at the highest fluence of 105, 91, and 81 displacement per atom, respectively, were studied. Transmission electron microscopy (TEM) and high-resolution TEM (HRTEM) analyses show that the initial microstructures and phase composition of all three alloys are retained after ion irradiation and no phase decomposition is observed. Furthermore, it is demonstrated that the disordered face-centered cubic (FCC) and disordered body-centered cubic (BCC) phases show much less defect cluster formation and structural damage than the NiAl-type ordered B2 phase. This effect is explained by higher entropy of mixing, higher defect formation/migration energies, substantially lower thermal conductivity, and higher atomic level stress in the disordered phases.

  9. Density decrease in vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Galvin, T.M.; Smith, D.L.

    1996-04-01

    Combined effects of dynamically charged helium and neutron damage on density decrease (swelling) of V-4Cr-4Ti, V-5Ti, V-3Ti-1Si, and V-8Cr-6Ti alloys have been determined after irradiation to 18-31 dpa at 425-600{degrees}C in the Dynamic helium Charging Experiment (DHCE). To ensure better accuracy in density measurement, broken pieces of tensile specimens {approx} 10 times heavier than a transmission electron microscopy (TEM) disk were used. Density increases of the four alloys irradiated in the DHCE were <0.5%. This small change seems to be consistent with the negligible number density of microcavities characterized by TEM. Most of the dynamically produced helium atoms seem to have been trapped in the grain matrix without significant cavity nucleation or growth.

  10. Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1997-04-01

    Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

  11. Neutron irradiation effects on the microstructural development of tungsten and tungsten alloys

    NASA Astrophysics Data System (ADS)

    Hasegawa, Akira; Fukuda, Makoto; Yabuuchi, Kiyohiro; Nogami, Shuhei

    2016-04-01

    Data on the microstructural development of tungsten (W) and tungsten rhenium (Re) alloys were obtained after neutron irradiation at 400-800 °C in the Japan Materials Testing Reactor (JMTR), the experimental fast test reactor Joyo, and the High Flux Isotope Reactor (HFIR) for irradiation damage levels in the range of 0.09-1.54 displacement per atom (dpa). Microstructural observations showed that a small amount of Re (3-5%) in W-Re alloys is effective in suppressing void formation. In W-Re alloys with Re concentrations greater than 10%, acicular precipitates are the primary structural defects. In the HFIR-irradiated specimen, in which a large amount of Re was expected to be produced by the nuclear transmutation of W to Re because of the reactor's high thermal neutron flux, voids were not observed even in pure W. The synergistic effects of displacement damage and solid transmutation elements on microstructural development are discussed, and the microstructural development of tungsten materials utilized in fusion reactors is predicted.

  12. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  13. Effects of neutron irradiation on mechanical properties and microstructures of dispersion and precipitation hardened copper alloys

    NASA Astrophysics Data System (ADS)

    Singh, B. N.; Edwards, D. J.; Toft, P.

    1996-11-01

    Tensile specimens of Cusbnd Al2O3, CuCrZr and CuNiBe alloys were irradiated with fission neutrons to fluences of 5 × 1022, 5 × 1023and1 × 1024n/m2 (E > 1MeV) at 47°C. Tensile properties and Vickers hardness were determined at 22°C. Microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope and the fractured surfaces were investigated in a scanning electron microscope. The most significant effect of irradiation is a drastic decrease in the ductility of copper alloys at a dose level as low as 0.2 dpa. The loss of ductility appears to be related to the intrinsic hardness of the grain interior and not to the grain boundary embrittlement. It is suggested that the irradiation-induced hardening and the lack of uniform elongation may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) and/or impurity atoms.

  14. Ar irradiated Cr rich Ni alloy studied using positron annihilation spectroscopy

    NASA Astrophysics Data System (ADS)

    Saini, Sanjay; Menon, Ranjini; Sharma, S. K.; Srivastava, A. P.; Mukherjee, S.; Nabhiraj, P. Y.; Pujari, P. K.; Srivastava, D.; Dey, G. K.

    2016-10-01

    The present study focuses on understanding the effect of Ar ion irradiation at room temperature on Cr rich Ni-Cr alloy. The alloy is irradiated with Ar9+ ions (energy 315 keV) for total dose varying from 9.3 × 1014 to 2.3 × 1016 ion/cm2. The changes in the microstructure of the irradiated samples have been characterized by depth dependent Doppler broadening of annihilation radiation (DBAR) measurements using a slow positron beam facility. The variation in S-E profiles as a function of total dose corroborated with S-W curves indicates that the type of defects is also varied with the increase in total dose. The S-E profiles have been fitted using variable energy positron fit (VEPFIT) program considering a three layer structure for the irradiated samples. Estimated displacement damage profile as a function of increasing dose has been analyzed and a possible mechanism has been attributed to explain the observations made from S-parameter variation.

  15. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    NASA Astrophysics Data System (ADS)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  16. Effects of ultraviolet irradiation on bonding strength between Co-Cr alloy and citric acid-crosslinked gelatin matrix.

    PubMed

    Inoue, Motoki; Sasaki, Makoto; Katada, Yasuyuki; Taguchi, Tetsushi

    2014-02-01

    Novel techniques for creating a strong bond between polymeric matrices and biometals are required. We immobilized polymeric matrices on the surface of biometal for drug-eluting stents through covalent bond. We performed to improve the bonding strength between a cobalt-chromium alloy and a citric acid-crosslinked gelatin matrix by ultraviolet irradiation on the surface of cobalt-chromium alloy. The ultraviolet irradiation effectively generated hydroxyl groups on the surface of the alloy. The bonding strength between the gelatin matrix and the alloy before ultraviolet irradiation was 0.38 ± 0.02 MPa, whereas it increased to 0.48 ± 0.02 MPa after ultraviolet irradiation. Surface analysis showed that the citric acid derivatives occurred on the surface of the cobalt-chromium alloy through ester bond. Therefore, ester bond formation between the citric acid derivatives active esters and the hydroxyl groups on the cobalt-chromium alloy contributed to the enhanced bonding strength. Ultraviolet irradiation and subsequent immobilization of a gelatin matrix using citric acid derivatives is thus an effective way to functionalize biometal surfaces.

  17. Properties of V-(8-9)Cr-(5-6)Ti alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    1996-10-01

    In the Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in vanadium alloy specimens by the decay of tritium during irradiation to 18-31 dpa at 425-600{degrees}C in lithium-filled capsules in the Fast Flux Test Facility. This report presents results of postirradiation tests of tensile properties and density change in V-8Cr-6Ti and V-9Cr-5Ti. Compared to tensile properties of the alloys irradiated in the non-DHCE (helium generation negligible), the effect of helium on tensile strength and ductility of V-8Cr-6Ti and V-9Cr-5Ti was insignificant after irradiation and testing at 420, 500, and 600{degrees}C. Both alloys retained a total elongation of >11 % at these temperatures. Density change was <0.48% for both alloys.

  18. NANOPARTICLES: Formation of the alloy of Au and Ag nanoparticles upon laser irradiation of the mixture of their colloidal solutions

    NASA Astrophysics Data System (ADS)

    Izgaliev, Andrei T.; Simakin, Aleksandr V.; Shafeev, Georgii A.

    2004-01-01

    The formation dynamics of the alloy of gold and silver nanoparticles is studied upon laser irradiation of the mixture of these nanoparticles and factors affecting the alloy formation are determined. Individual nanoparticles are obtained by ablation of the corresponding metals in water or ethanol by copper vapour laser radiation at a wavelength of 510.6 nm close to the maximum of the plasmon resonance of gold particles at 518 nm. The intermediate phase of the alloy characterised by an anomalous red shift of the absorption spectrum is found for the first time. The dependences of the absorption spectrum of the alloy of colloidal particles of these metals and their morphology on the irradiation time are obtained. It is found that the rate of the alloy formation depends on the concentrations of nanoparticles and surfactants in the mixture.

  19. A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

    2016-05-01

    Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

  20. Effect of He+ irradiation on Fe-Cr alloys: Mössbauer-effect study

    NASA Astrophysics Data System (ADS)

    Dubiel, S. M.; Cieślak, J.; Reuther, H.

    2013-03-01

    Effect of He ions irradiation on three model Fe100-xCrx alloys (x = 5.8, 10.75 and 15.15) was investigated with the conversion electron Mössbauer spectroscopy. The study of the alloys irradiated with 25 keV ions revealed that the strongest effect occured in the Fe84.85Cr15.15 sample where an inversion of a short-range-order (SRO) parameter was found. Consequently, the investigation of the influence of the irradiation dose, D, was carried out on the chromium - most concentrated sample showing that the average hyperfine field, , the average angle between the normal to the sample's surface and the magnetization vector, <θ>, as well as the actual distribution of Fe/Cr atoms, as expressed by SRO parameters, strongly depend on D. In particular: (a) increases with D, and its maximum increase corresponds to a decrease of Cr content within the two-shell volume around the probe 57Fe nuclei by ˜2.3 at%, <⊝> decreases by ˜13° at maximum, (c) SRO-parameter averaged over the two-shell volume increases with D from weakly negative value (indicative of Cr atoms ordering) to weakly positive value (indicative of Cr atoms clustering). The inversion takes place at D ≈ 7 dpa.

  1. Tensile and impact properties of vanadium-base alloys irradiated at low temperatures in the ATR-A1 experiment

    SciTech Connect

    Tsai, H.; Nowicki, L.J.; Billone, M.C.; Chung, H.M.; Smith, D.L.

    1998-03-01

    Subsize tensile and Charpy specimens made from several V-(4-5)Cr-(4-5)Ti alloys were irradiated in the ATR-A1 experiment to study the effects of low-temperature irradiation on mechanical properties. These specimens were contained in lithium-bonded subcapsules and irradiated at temperatures between {approx}200 and 300 C. Peak neutron damage was {approx}4.7 dpa. Postirradiation testing of these specimens has begun. Preliminary results from a limited number of specimens indicate a significant loss of work-hardening capability and dynamic toughness due to the irradiation. These results are consistent with data from previous low-temperature neutron irradiation experiments on these alloys.

  2. Phase stability of decomposed Ni-Al alloys under ion irradiation

    SciTech Connect

    Schmitz, G.; Ewert, J. C.; Harbsmeier, F.; Uhrmacher, M.; Haider, F.

    2001-06-01

    The disordering of a decomposed Ni-12at.%-Al alloy under irradiation with 300-keV Ni{sup +} ions is studied in direct comparison to homogeneous Ni{sub 3}Al at various temperatures. For that, the degree of long-range order is quantitatively determined by electron diffractometry. Heterogeneous alloys disorder significantly faster than the homogeneous intermetallic. In addition, their disordering rate depends on the precipitate size of the microstructure. At 823 K a steady state with an intermediate degree of order is reached. The experimental results are compared with model calculation based on a Cahn-Hilliard-type continuum model. At higher temperatures the model yields a correct description of the experiments. However, discrepancies at room temperature suggest a direct coupling of local composition and degree of order during the collision cascade. This interpretation is further supported by Monte Carlo simulations of overlapping cascade events.

  3. Atomic kinetic Monte Carlo modeling of multi-component Fe dilute alloys under irradiation

    NASA Astrophysics Data System (ADS)

    Domain, C.; Becquart, C. S.

    2014-06-01

    The ageing of pressure vessel steels under radiation has been correlated with the formation of more or less dilute solute clusters which are investigated in this work using a multiscale approach based on ab initio and atomistic kinetic Monte Carlo (AKMC) simulations. The microstructure evolution of Fe alloys is modeled by AKMC on a lattice, using pair interactions adjusted on DFT calculations. Several substitutional elements (Cu, Ni, Mn, Si, P) and foreign interstitials (C, N) are taken into account to describe the alloy. The point defect created by the irradiation, i.e. the vacancies and self interstitials have a tendency to form clusters. The evolution of these clusters is governed by the migration energy of the individual point defects which is very heavy in terms of computing time due to the large number of AKMC steps required. The structure of all the possible objects that can form is complex and some optimized and accelerated methods will be presented.

  4. Microstructure and mechanical properties of neutron irradiated V-20Ti alloy

    NASA Astrophysics Data System (ADS)

    Higashiguchi, Y.; Kayano, H.; Morozumi, S.

    1985-08-01

    Neutron irradiation damage and its effects on mechanical properties were studied for V-20Ti alloy with three different fluences: 5 × 10 24, 2 × 10 24 and 1× 10 23 n m -2 ( E > l MeV) , named A-, B- and C-specimens, respectively. Radiation-induced precipitates (RIP) were observed in A and B specimens and no voids in all specimens. It was clear from the detailed investigation that the precipitates are circular and planar coherent particles lying on (100) planes and have the fee structures. The mechanisms for suppression of void formation and for softening the irradiation effects on the mechanical properties were suggested based on the relation with radiation-induced precipitates (RIP) which were considered as a non-stoichiometric suboxide and contained many vacancies at lattice sites.

  5. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L.; Matsui, H.

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  6. Observations of void swelling in selected austenitic alloys during ion irradiation under a rising temperature ramp

    NASA Astrophysics Data System (ADS)

    Mazey, D. J.; Williams, T. M.; Bolster, D. E. J.

    1988-07-01

    Results are given of a TEM study of the void swelling behaviour of CW AISI 321 (En58B), CW FV548 stainless steel and STA Nimonic PE16 alloy during 46 MeV nickel ion irradiation under a near-linear rising temperature ramp. The dose range for the ramp was 0-75 dpa and four temperature ranges were investigated using 50 ° C intervals from 500-550 ° C, 525-575 ° C, 575-625 ° C and 625-675 ° C. Swelling in CW 321 was found to be no greater than that observed after isothermal irradiation to 75 dpa at temperatures corresponding to the end-of-ramp temperatures. This was not the case for CW FV548 where the swelling for the temperature ranges of 525-575 ° C and 575-625 ° C was slightly greater than that observed under isothermal irradiation to 75 dpa. Swelling in STA PE16 in the lower temperature ranges of 500-550 ° C and 525-575°C was much higher than corresponding isothermal values but in the 600-700 ° C temperature range the swelling was close to isothermal values. The results on PE16 indicate that a steady increase in irradiation temperature from just below or just above the low temperature cut-off into the void swelling region produces a significant increase in swelling with respect to that expected under isothermal conditions. Possible interpretations of the observed swelling behaviour in these alloys are discussed.

  7. Tensile properties of V-(4-15)Cr-5Ti alloys irradiated at 400{degrees}C in the HFIR

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    1996-10-01

    V-(4-15)Cr-5Ti alloys were irradiated in a helium environment to {approx}10 dpa at {approx}400{degrees}C in the High Flux Isotope Reactor (HFIR). This report presents results of postirradiation tests of tensile properties of V-4Cr-4Ti, V-8Cr-6Ti, V-10Cr-5Ti, and V-15Cr-5Ti. Despite concerns on the effects of transmutation of vanadium to Cr and impurity pickup from the helium environment, all of the alloys exhibited ductile tensile behavior. However, the alloys exhibited ductilities somewhat lower than those of the specimens irradiated to a similar dose and at a similar temperature in an Li environment in fast reactors. Uniform plastic strain in the V-Cr-(4-5)Ti alloys decreased monotonically with increasing Cr content.

  8. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    SciTech Connect

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.; Rowcliffe, A.F.

    1996-04-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of {approx}0.5 to 5 dpa and temperatures between {approx}80 and 400{degrees}C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200{degrees}C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of {approx}1.2n{Omega}m for pure copper and {approx}1.6n{Omega}m for dispersion strengthened copper irradiated at {approx}100{degrees}C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than {approx}5 10{sup 24} n.m{sup 2}. The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above {approx}300{degrees}C.

  9. Subtask 12F4: Effects of neutron irradiation on the impact properties and fracture behavior of vanadium-base alloys

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1995-03-01

    Up-to-date results on the effects of neutron irradiation on the impact properties and fracture behavior of V, V-Ti, V-Cr-Ti and V-Ti-Si alloys are presented in this paper, with an emphasis on the behavior of the U.S. reference alloys V-4Cr-4Ti containing 500-1000 wppm Si. Database on impact energy and cluctile-brittle transition temperature (DBTT) has been established from Charpy impact tests of one-third-size specimens irradiated at 420{degrees}C-600{degrees}C up to {approx}50 dpa in lithium environment in fast fission reactors. To supplement the Charpy impact tests fracture behavior was also characterized by quantitative SEM fractography on miniature tensile and disk specimens that were irradiated to similar conditions and fractured at -196{degrees}C to 200{degrees}C by multiple bending. For similar irradiation conditions irradiation-induced increase in DBTT was influenced most significantly by Cr content, indicating that irradiation-induced clustering of Cr atoms takes place in high-Cr (Cr {ge} 7 wt.%) alloys. When combined contents of Cr and Ti were {le}10 wt.%, effects of neutron irradiation on impact properties and fracture behavior were negligible. For example, from the Charpy-impact and multiple-bend tests there was no indication of irradiation-induced embrittlement for V-5Ti, V-3Ti-1Si and the U.S. reference alloy V-4Cr-4Ti after irradiation to {approx}34 dpa at 420{degrees}C to 600{degrees}C, and only ductile fracture was observed for temperatures as low as -196{degrees}C. 14 refs., 8 figs., 1 tab.

  10. Microstructural evolution of Fesbnd 22%Cr model alloy under thermal ageing and ion irradiation conditions studied by atom probe tomography

    NASA Astrophysics Data System (ADS)

    Korchuganova, Olesya A.; Thuvander, Mattias; Aleev, Andrey A.; Rogozhkin, Sergey V.; Boll, Torben; Kulevoy, Timur V.

    2016-08-01

    Nanostructure evolution during ion irradiation of two thermally aged binary Fee22Cr alloys has been investigated using atom probe tomography. Specimens aged at 500 °C for 50 and 200 h were irradiated by 5.6 MeV Fe ions at room temperature up to fluences of 0.3 × 1015 ions/cm2 and 1 × 1015 ions/cm2. The effect of irradiation on the material nanostructure was examined at a depth of 1 μm from the irradiated surface. The analysis of Cr radial concentration functions reveals that dense α‧-phase precipitates in the 200 h aged alloy become diffuse and thereby larger when subjected to irradiation. On the other hand, less Cr-enriched precipitates in the alloy aged for 50 h are less affected. The CreCr pair correlation function analysis shows that matrix inhomogeneity decreases under irradiation. Irradiation leads to a decrease in the number density of diffuse clusters, whereas in the case of well-developed precipitates it remains unchanged.

  11. Phase decomposition of AuFe alloy nanoparticles embedded in silica matrix under swift heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Pannu, Compesh; Bala, Manju; Singh, U. B.; Srivastava, S. K.; Kabiraj, D.; Avasthi, D. K.

    2016-07-01

    AuFe alloy nanoparticles embedded in silica matrix are synthesized using atom beam sputtering technique and subsequently irradiated with 100 MeV Au ions at various fluences ranging from 1 × 1013 to 6 × 1013 ions/cm2. The X-ray diffraction, absorption spectroscopy, X-ray photo electron spectroscopy and transmission electron microscopy results show that swift heavy ion irradiation leads to decomposition of AuFe alloy nanoparticles from surface region and subsequent reprecipitation of Au and Fe nanoparticles occur. The process of phase decomposition and reprecipitation of individual element nanoparticles is explained on the basis of inelastic thermal spike model.

  12. Void structure and density change of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Gazda, J.

    1995-04-01

    The objective of this work is to determine void structure, distribution, and density changes of several promising vanadium-base alloys irradiated in the Dynamic Helium Charging Experiment (DHCE). Combined effects of dynamically charged helium and neutron damage on density change, void distribution, and microstructural evolution of V-4Cr-4Ti alloy have been determined after irradiation to 18-31 dpa at 425-600{degree}C in the DHCE, and the results compared with those from a non-DHCE in which helium generation was negligible.

  13. Ion irradiation studies on the void swelling behavior of a titanium modified D9 alloy

    NASA Astrophysics Data System (ADS)

    Balaji, S.; Mohan, Sruthi; Amirthapandian, S.; Chinnathambi, S.; David, C.; Panigrahi, B. K.

    2015-12-01

    The sensitivity of Positron Annihilation Spectroscopy (PAS) for probing vacancy defects and their environment is well known. Its applicability in determination of swelling and the peak swelling temperature was put to test in our earlier work on ion irradiated D9 alloys [1]. Upon comparison with the peak swelling temperature determined by conventional step height measurements it was found that the peak swelling temperature determined using PAS was 50 K higher. It was conjectured that the positrons trapping in the irradiation induced TiC precipitation could have caused the shift. In the present work, D9 alloys have been implanted with 100 appm helium ions and subsequently implanted with 2.5 MeV Ni ions up to peak damage of 100 dpa. The nickel implantations have been carried out through a range of temperatures between 450 °C and 650 °C. The evolution of cavities and TiC precipitates at various temperatures has been followed by TEM and this report provides an experimental verification of the conjecture.

  14. Nano-Scale Fission Product Phases in an Irradiated U-7Mo Alloy Nuclear Fuel

    SciTech Connect

    Dennis Keiser, Jr.; Brandon Miller; James Madden; Jan-Fong Jue; Jian Gan

    2014-09-01

    Irradiated nuclear fuel is a very difficult material to characterize. Due to the large radiation fields associated with these materials, they are hard to handle and typically have to be contained in large hot cells. Even the equipment used for performing characterization is housed in hot cells or shielded glove boxes. The result is not only a limitation in the techniques that can be employed for characterization, but also a limitation in the size of features that can be resolved The most standard characterization techniques include light optical metallography (WM), scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). These techniques are applied to samples that are typically prepared using grinding and polishing approaches that will always generate some mechanical damage on the sample surface. As a result, when performing SEM analysis, for example, the analysis is limited by the quality of the sample surface that can be prepared. However, a new approach for characterizing irradiated nuclear fuel has recently been developed at the Idaho National Laboratory (INL) in Idaho Falls, Idaho. It allows for a dramatic improvement in the quality of characterization that can be performed when using an instrument like an SEM. This new approach uses a dual-beam scanning microscope, where one of the beams isa focused ion beam (FIB), which can be used to generate specimens of irradiated fuel (-10µm x 10µm) for microstructural characterization, and the other beam is the electron beam of an SEM. One significant benefit of this approach is that the specimen surface being characterized has received much less damage (and smearing) than is caused by the more traditional approaches, which enables the imaging of nanometer­ sized microstructural features in the SEM. The process details are for an irradiated low-enriched uranium (LEU) U-Mo alloy fuel Another type of irradiated fuel that has been characterized using this technique is a mixed oxide fuel.

  15. Irradiation-induced precipitation and mechanical properties of vanadium alloys at <430 C

    SciTech Connect

    Chung, H.M.; Gazda, J.; Smith, D.L.

    1998-09-01

    Recent attention to V-base alloys has focused on the effect of low-temperature (<430 C) irradiation on tensile and impact properties of V-4Cr-4Ti. In previous studies, dislocation channeling, which causes flow localization and severe loss of work-hardening capability, has been attributed to dense, irradiation-induced precipitation of very fine particles. However, efforts to identify the precipitates were unsuccessful until now. In this study, analysis by transmission electron microscopy (TEM) was conducted on unalloyed V, V-5Ti, V-3Ti-1Si, and V-4Cr-4Ti specimens that were irradiated at <430 C in conventional and dynamic helium charging experiments. By means of dark-field imaging and selected-area-diffraction analysis, the characteristic precipitates were identified to be (V,Ti{sub 1{minus}x})(C,O,N). In V-3Ti-1Si, precipitation of (V,Ti{sub 1{minus}x})(C,O,N) was negligible at <430 C, and as a result, dislocation channeling did not occur and work-hardening capability was high.

  16. Effect of neutron irradiation and postradiation annealing on the microstructure and properties of an Al-Mg-Si alloy

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.; Tsai, K. V.; Rofman, O. V.; Sil'nyagina, N. S.

    2016-09-01

    The effect of long-term neutron irradiation and postradiation thermal-induced aging on the microstructure and mechanical properties of an aluminum-based reactor Al-Mg-Si alloy grade SAV-1 has been studied. The material under study is the shell of an automatic fine-control rod used to control the reactivity of the core of a VVR-K research reactor. Successive 1-h annealings of specimens of the SAV-1 alloy irradiated to doses of 0.001 and 5 dpa in the temperature range of 100-550°C have been carried out. The evolution of the fine structure of the material and changes in its mechanical characteristics have been studied. The phenomenon of the acceleration of the aging of the SAV-1 alloy under the effect of a high neutron fluence at an irradiation temperature of 80°C has been observed, which involves the formation of numerous lineage (stitch) Guinier-Preston zones in the alloy. It has been shown that the strength characteristics of the SAV-1 alloy depend significantly on the degree of its radiation- and thermal-induced aging.

  17. Effect of fast neutron irradiation on fatigue-crack growth behavior of three nickel-base alloys

    SciTech Connect

    James, L.A.

    1981-04-01

    Nickel-base alloys are often employed in reactor structural applications where high strength, resistance to corrosion, or swelling resistance are important considerations. In this study, fatigue-crack growth rate tests were conducted at 427/degree/C on Inconel 600, Inconel X-750, and Incoloy 800 irradiated in the Experimental Breeder Reactor II to total fluences ranging between 2.5 and 6.0*10/sup 22/ n/cm/sup 2/. Following irradiation, minor increases were noted in the crack growth rates of Inconel 600, minor decreases in the growth rates of Inconel X-750, and no irradiation effect in the cracking behavior of Incoloy 800. 14 refs.

  18. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  19. A generalized Ising model for studying alloy evolution under irradiation and its use in kinetic Monte Carlo simulations.

    PubMed

    Huang, Chen-Hsi; Marian, Jaime

    2016-10-26

    We derive an Ising Hamiltonian for kinetic simulations involving interstitial and vacancy defects in binary alloys. Our model, which we term 'ABVI', incorporates solute transport by both interstitial defects and vacancies into a mathematically-consistent framework, and thus represents a generalization to the widely-used ABV model for alloy evolution simulations. The Hamiltonian captures the three possible interstitial configurations in a binary alloy: A-A, A-B, and B-B, which makes it particularly useful for irradiation damage simulations. All the constants of the Hamiltonian are expressed in terms of bond energies that can be computed using first-principles calculations. We implement our ABVI model in kinetic Monte Carlo simulations and perform a verification exercise by comparing our results to published irradiation damage simulations in simple binary systems with Frenkel pair defect production and several microstructural scenarios, with matching agreement found. PMID:27541350

  20. Formation of Pt-Zn Alloy Nanoparticles by Electron-Beam Irradiation of Wurtzite ZnO in the TEM

    NASA Astrophysics Data System (ADS)

    Lee, Sung Bo; Park, Jucheol; van Aken, Peter A.

    2016-07-01

    As is well documented, platinum nanoparticles, promising for catalysts for fuel cells, exhibit better catalytic activities, when alloyed with Zn. Pre-existing syntheses of Pt-Zn alloy catalysts are composed of a number of complex steps. In this study, we have demonstrated that nanoparticles of Pt-Zn alloys are simply generated by electron-beam irradiation in a transmission electron microscope of a wurtzite ZnO single-crystal specimen. The initial ZnO specimen is considered to have been contaminated by Pt during specimen preparation by focused ion beam milling. The formation of the nanoparticle is explained within the framework of ionization damage (radiolysis) by electron-beam irradiation and accompanying electrostatic charging.

  1. Formation of Pt-Zn Alloy Nanoparticles by Electron-Beam Irradiation of Wurtzite ZnO in the TEM.

    PubMed

    Lee, Sung Bo; Park, Jucheol; van Aken, Peter A

    2016-12-01

    As is well documented, platinum nanoparticles, promising for catalysts for fuel cells, exhibit better catalytic activities, when alloyed with Zn. Pre-existing syntheses of Pt-Zn alloy catalysts are composed of a number of complex steps. In this study, we have demonstrated that nanoparticles of Pt-Zn alloys are simply generated by electron-beam irradiation in a transmission electron microscope of a wurtzite ZnO single-crystal specimen. The initial ZnO specimen is considered to have been contaminated by Pt during specimen preparation by focused ion beam milling. The formation of the nanoparticle is explained within the framework of ionization damage (radiolysis) by electron-beam irradiation and accompanying electrostatic charging. PMID:27440080

  2. A generalized Ising model for studying alloy evolution under irradiation and its use in kinetic Monte Carlo simulations

    NASA Astrophysics Data System (ADS)

    Huang, Chen-Hsi; Marian, Jaime

    2016-10-01

    We derive an Ising Hamiltonian for kinetic simulations involving interstitial and vacancy defects in binary alloys. Our model, which we term ‘ABVI’, incorporates solute transport by both interstitial defects and vacancies into a mathematically-consistent framework, and thus represents a generalization to the widely-used ABV model for alloy evolution simulations. The Hamiltonian captures the three possible interstitial configurations in a binary alloy: A-A, A-B, and B-B, which makes it particularly useful for irradiation damage simulations. All the constants of the Hamiltonian are expressed in terms of bond energies that can be computed using first-principles calculations. We implement our ABVI model in kinetic Monte Carlo simulations and perform a verification exercise by comparing our results to published irradiation damage simulations in simple binary systems with Frenkel pair defect production and several microstructural scenarios, with matching agreement found.

  3. Morphological control and evolution of octahedral and truncated trisoctahedral Pt-Au alloy nanocrystals under microwave irradiation

    NASA Astrophysics Data System (ADS)

    Dai, Lei; Zhao, Yanxi; Chi, Quan; Liu, Hanfan; Li, Jinlin; Huang, Tao

    2014-08-01

    Uniform and well-defined truncated trisoctahedral and octahedral Pt-Au alloy nanocrystals were fabricated by co-reducing H2PtCl6-HAuCl4 with tetraethylene glycol (TEG) under microwave irradiation for only 140 s. Iodide ions were critical to the morphological control and evolution of Pt-Au alloy nanostructures. The as-prepared Pt-Au alloy nanocrystals exhibited efficient electrocatalytic activities.Uniform and well-defined truncated trisoctahedral and octahedral Pt-Au alloy nanocrystals were fabricated by co-reducing H2PtCl6-HAuCl4 with tetraethylene glycol (TEG) under microwave irradiation for only 140 s. Iodide ions were critical to the morphological control and evolution of Pt-Au alloy nanostructures. The as-prepared Pt-Au alloy nanocrystals exhibited efficient electrocatalytic activities. Electronic supplementary information (ESI) available: Experimental details; SEM, TEM and HAADF-STEM images, UV-vis absorbance spectra, XRD. See DOI: 10.1039/c4nr01864h

  4. In situ observation of defect annihilation in Kr ion-irradiated bulk Fe/amorphous-Fe 2 Zr nanocomposite alloy

    DOE PAGES

    Yu, K. Y.; Fan, Z.; Chen, Y.; Song, M.; Liu, Y.; Wang, H.; Kirk, M. A.; Li, M.; Zhang, X.

    2014-08-26

    Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe96Zr4 nanocomposite alloy. Irradiation resulted in amorphization of Fe2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphous nanocomposites.

  5. Microstructural stability and mechanical behavior of FeNiMnCr high entropy alloy under ion irradiation

    DOE PAGES

    Leonard, Keith J.; Bei, Hongbin; Zinkle, Steven J.; Kiran Kumar, N. A. P.; Li, C.

    2016-05-13

    In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent mechanical properties and good corrosion resistance, making them potential candidates for high temperature fission and fusion structural applications. However there is very little known about their radiation resistance, particularly at elevated temperatures relevant for energy applications. In the present study, a single phase (face centered cubic) concentrated solid solution alloy of composition 27%Fe-28%Ni-27%Mn-18%Cr was irradiated with 3 or 5.8 MeV Ni ions at temperatures ranging from room temperature to 700 °C and midrange doses from 0.03 to 10 displacements per atom (dpa). Transmission electron microscopymore » (TEM), scanning transmission electron microscopy with energy dispersive x-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterize the radiation defects and microstructural changes. Irradiation at higher temperatures showed evidence of relatively sluggish solute diffusion with limited solute depletion or enrichment at grain boundaries. The main microstructural feature at all temperatures was high-density small dislocation loops. Voids were not observed at any irradiation condition. Nano-indentation tests on specimens irradiated at room temperature showed a rapid increase in hardness ~35% and ~80% higher than the unirradiated value at 0.03 and 0.3 dpa midrange doses, respectively. The irradiation-induced hardening was less pronounced for 500 °C irradiations (<20% increase after 3 dpa). Overall, the examined HEA material exhibits superior radiation resistance compared to conventional single phase Fe-Cr-Ni austenitic alloys such as stainless steels. Furthermore, the present study provides insight on the fundamental irradiation behavior of a single phase HEA material over a broad range of irradiation temperatures.« less

  6. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  7. Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

    SciTech Connect

    Edwards, D.J.; Singh, B.N.; Toft, P.; Eldrup, M.

    1997-04-01

    Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

  8. Microstructure of V-4Cr-4Ti alloy after low-temperature irradiation by ions and neutrons

    SciTech Connect

    Gazda, J.; Meshii, M.; Chung, H.M.

    1998-03-01

    Mechanical properties of V-4Cr-4Ti alloy were investigated after low-temperature (<420 C) irradiation. The effects of fast neutrons at 390 C were investigated by irradiation to {approx}4 dpa in the X530 experiment in the EBR-II reactor; these tests were complemented by irradiation with single (4.5-MeV Ni{sup ++}) and dual ion beams (350-keV He{sup +} simultaneously with 4.5-MeV Ni{sup ++}). TEM observations showed the formation of a high density of point-defect clusters and dislocation loops (<30 nm diameter) distributed uniformly in the specimens. Mechanical-property testing showed embrittlement of the alloy. TEM investigations of deformed microstructures were used to determine the causes of embrittlement and yielded observation of dislocation channels propagating through the undeformed matrix. Channels are the sole slip paths and cause early onset of necking and loss of work-hardening in this alloy. Based on a review of the available literature, suggestions are made for further research of slip localization in V-base alloys.

  9. Nanocavity formation and hardness increase by dual ion beam irradiation of oxide dispersion strengthened FeCrAl alloy

    NASA Astrophysics Data System (ADS)

    Kögler, R.; Anwand, W.; Richter, A.; Butterling, M.; Ou, Xin; Wagner, A.; Chen, C.-L.

    2012-08-01

    Open volume defects generated by ion implantation into oxide dispersion strengthened (ODS) alloy and the related hardness were investigated by positron annihilation spectroscopy and nanoindentation measurements, respectively. Synchronized dual beam implantation of Fe and He ions was performed at room temperature and at moderately enhanced temperature of 300 °C. For room temperature implantation a significant hardness increase after irradiation is observed which is more distinctive in heat treated than in as-received ODS alloy. There is also a difference between the simultaneous and sequential implantation mode as the hardening effect for the simultaneously implanted ODS alloy is stronger than for sequential implantation. The comparison of hardness profiles and of the corresponding open volume profiles shows a qualitative agreement between the open volume defects generated on the nanoscopic scale and the macroscopic hardness characteristics. Open volume defects are drastically reduced for performing the simultaneous dual beam irradiation at 300 °C which is a more realistic temperature under application aspects. Few remaining defects are clusters of 3-4 vacancies in connection with Y oxide nanoparticles. These defects completely disappear in a shallow layer at the surface. The results are in agreement with hardness measurements showing little hardness increase after irradiation at 300 °C. Suitable characteristics of ODS alloy for nuclear applications and the close correlation between He-related open volume defects and the hardness characteristics are verified.

  10. A replica technique for extracting precipitates from neutron-irradiated or thermal-aged vanadium alloys for TEM analysis

    NASA Astrophysics Data System (ADS)

    Fukumoto, K.; Iwasaki, M.

    2014-06-01

    A carbon replica technique has been developed to extract precipitates from vanadium alloys. Using this technique, precipitation phases can be extracted from neutron-irradiated or thermal-aged V-4Cr-4Ti alloys. Precipitate identification using EDS X-ray analysis and electron diffraction was facilitated. Only NaCl type of Ti(OCN) precipitate was formed in the thermal-aged V-4Cr-4Ti alloys at 600 °C for 20 h and cation sub-lattice was only occupied by Ti atoms. However, the thin plate of precipitates with NaCl type of crystallographic structure could be seen in the V-4Cr-4Ti alloys irradiated at 593 °C in the JOYO fast reactor. The precipitate contained chromium and vanadium atoms on the cation sub-lattice as well as titanium atoms. It is considered that the phase of MX type (M = Ti, V, Cr and X = O, N, C) is a metastable phase under neutron irradiation.

  11. Temperature and thermal stress fields during the pulse train of long-pulse laser irradiating aluminium alloy plate

    NASA Astrophysics Data System (ADS)

    Zhang, Wei; Jin, Guangyong; Gu, Xiu-ying

    2014-12-01

    Based on Von Mises yield criterion and elasto-plastic constitutive equations, an axisymmetric finite element model of a Gaussian laser beam irradiating a metal substrate was established. In the model of finite element, the finite difference hybrid algorithm is used to solve the problem of transient temperature field and stress field. Taking nonlinear thermal and mechanical properties into account, transient distributions of temperature field and stress fields generated by the pulse train of long-pulse laser in a piece of aluminium alloy plate were computed by the model. Moreover,distributions as well as histories of temperature and stress fields were obtained. Finite element analysis software COMSOL is used to simulate the Temperature and thermal stress fields during the pulse train of long-pulse laser irradiating 7A04 aluminium alloy plate. By the analysis of the results, it is found that Mises equivalent stress on the target surface distribute within the scope of the center of a certain radius. However, the stress is becoming smaller where far away from the center. Futhermore, the Mises equivalent stress almost does not effect on stress damage while the Mises equivalent stress is far less than the yield strength of aluminum alloy targets. Because of the good thermal conductivity of 7A04 aluminum alloy, thermal diffusion is extremely quick after laser irradiate. As a result, for the multi-pulsed laser, 7A04 aluminum alloy will not produce obvious temperature accumulation when the laser frequency is less than or equal to 10 Hz. The result of this paper provides theoretical foundation not only for research of theories of 7A04 aluminium alloy and its numerical simulation under laser radiation but also for long-pulse laser technology and widening its application scope.

  12. Mechanical properties of high-nickel alloys EP-753 and РЕ-16 after neutron irradiation to 54 dpa at 400-650 °С

    NASA Astrophysics Data System (ADS)

    Konobeev, Yu. V.; Porollo, S. I.; Ivanov, A. A.; Shulepin, S. V.; Budylkin, N. I.; Mironova, E. G.; Garner, F. A.

    2011-05-01

    Short-term mechanical properties and void swelling were investigated for high-nickel alloys РЕ-16 and three compositional variants of Russian alloy EP-753 and in various starting conditions after side-by-side irradiation in the BN-350 fast reactor at 400, 500, 600 and 650 °С to 54 dpa. For both alloys irradiation resulted in significant hardening and ductility reduction dependent on their chemical composition and initial heat treatment. At test temperatures equal to the irradiation values both alloys exhibited a high level of strength and satisfactory ductility. In the test temperature range of 550-650 °С the phenomenon of high-temperature irradiation embrittlement was observed.

  13. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation.

    SciTech Connect

    Long, Fei; Daymond, Mark R.; Yao, Zhongwen; Kirk, Marquis A.

    2015-03-14

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic < a > dislocations has been dynamically observed before and after irradiation at room temperature and 300 degrees C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by < a > dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, {01 (1) over bar1}< 0 (1) over bar 12 > twinning was identified in the irradiated sample deformed at 300 degrees C.

  14. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    SciTech Connect

    Rowcliffe, A.F.; Tsai, H.C.; Smith, D.L.

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  15. Effects of bonding bakeout thermal cycles on pre- and post irradiation microstructures, physical, and mechanical properties of copper alloys

    SciTech Connect

    Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J.

    1996-10-01

    At present, dispersion strengthened (DS) copper is being considered as the primary candidate material for the ITER first wall and divertor components. Recently, it was agreed among the ITER parties that a backup alloy should be selected from the two well known precipitation hardened copper alloys, CuCrZr and CuNiBe. It was therefore decided to carry out screening experiments to simulate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties, and electrical resistivity of CuCrZr and CuNiBe alloys. On the basis of the results of these experiments, one of the two alloys will be selected as a backup material. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime ageing, and bonding thermal cycle followed by reageing and the reactor bakeout treatment at 623K for 100 hours. Tensile specimens of the DS copper were also given the heat treatment corresponding to the bonding thermal cycle. A number of these heat treated specimens of CuCrZr, CuNiBe, and DS copper were neutron irradiated at 523K to a dose level of {approx}0.3 dpa (NRT) in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens with the various heat treatments were tensile tested at 532K. The dislocation, precipitate and void microstructures and electrical resistivity of these specimens were also determined. Results of these investigations will be reported and discussed in terms of thermal and irradiation stability of precipitates and irradiation-induced precipitation and recovery of dislocation microstructure. Results show that the bonding and bakeout thermal cycles are not likely to have any serious deleterious effects on the performance of these alloys. The CuNiBe alloys were found to be susceptible to radiation-induced embrittlement, however, the exact mechanism is not yet known. It is thought that radiation-induced precipitation and segregation of the beryllium may be responsible.

  16. The dependence of helium generation rate on nickel content of Fe-Cr-Ni alloys irradiated at high dpa levels in fast reactors

    SciTech Connect

    Garner, F.A.; Oliver, B.M.; Greenwood, L.R.

    1997-04-01

    With a few exceptions in the literature, it is generally accepted that it is nickel in Fe-Cr-Ni alloys that produces most of the transmutant helium and that the helium generation rate should scale linearly with the nickel content. Surprisingly, this assumption is based only on irradiations of pure nickel and has never been tested in an alloy series. There have also been no extensive tests of the predictions for helium production in alloys in various fast reactors spectra.

  17. Self-organization of Cu-based immiscible alloys under irradiation: An atom-probe tomography study

    NASA Astrophysics Data System (ADS)

    Stumphy, Brad D.

    The stability of materials subjected to prolonged irradiation has been a topic of renewed interest in recent years due to the projected growth of nuclear power as an alternative energy source. The irradiating particles impart energy into the material, thereby causing atomic displacements to occur. These displacements result in the creation of point defects and the random ballistic mixing of the atoms. Consequently, the material is driven away from its equilibrium structure. The supersaturation of defects can lead to the degradation of mechanical properties, but a high density of internal interfaces, which act as defect sinks, will suppress the supersaturation and long-range transport of defects. The microstructural evolution of the material is controlled by the ballistic mixing as well as the mobility of the point defects. In immiscible alloys, these two processes compete against one another, as the ballistic mixing acts to solutionize the alloy components, and the thermal diffusion of the large number of defects acts to phase separate the components. The work presented in this dissertation examines the effect of heavy-ion irradiation on immiscible, binary Cu-based alloys. Dilute alloys of Cu-Fe, Cu-V, and V-Cu have been subjected to irradiation, and atom-probe tomography has been utilized in order to better understand the complex nature of the response of these simple model systems to an irradiation environment. The results show that a steady-state, nano-scale patterning structure, with a high density of unsaturable defect sinks, can be maintained under prolonged irradiation. Additionally, precipitation from a supersaturated solid solution is shown to be a function of both the thermal diffusion and the ballistic mixing. Solvent-rich secondary precipitates, termed "cherry-pits," are observed inside of the solute-rich primary precipitates. Through a combination of simulation work and analyzing multiple alloys experimentally, it was determined that this cherry

  18. Void swelling in high purity FeCrNi and FeCrNiTi alloys irradiated in JOYO

    NASA Astrophysics Data System (ADS)

    Muroga, T.; Araki, K.; Miyamoto, Y.; Yoshida, N.

    1988-07-01

    Microstructures have been observed in Fe-13Cr-14Ni and Fe-13Cr-14Ni-0.12Ti alloys irradiated in JOYO (Japanese Fast Experimental Reactor) at 400, 500 and 600 °C to the fluence of 0.079, 0.81 and 6.2 × 10 25n/ m2 ( E > 0.1 MeV). In the Fe-13Cr-14Ni alloy, voids are observed in all cases. The dose dependence of swelling seems to obey the kinetics of linear increase with or without initial short transient. On the other hand, remarkable swelling suppression effects are observed in the Fe-13Cr-14Ni-0.12Ti alloy. The detailed microstructural observation suggests the titanium addition effects suppress the void nucleation in the matrix by gettering impurities and obstructing dislocation climb by precipitate decoration on dislocation lines.

  19. Microchemical effects in irradiated Fe-Cr alloys as revealed by atomistic simulation

    NASA Astrophysics Data System (ADS)

    Malerba, L.; Bonny, G.; Terentyev, D.; Zhurkin, E. E.; Hou, M.; Vörtler, K.; Nordlund, K.

    2013-11-01

    Neutron irradiation produces evolving nanostructural defects in materials, that affect their macroscopic properties. Defect production and evolution is expected to be influenced by the chemical composition of the material. In turn, the accumulation of defects in the material results in microchemical changes, which may induce further changes in macroscopic properties. In this work we review the results of recent atomic-level simulations conducted in Fe-Cr alloys, as model materials for high-Cr ferritic-martensitic steels, to address the following questions: 1. Is the primary damage produced in displacement cascades influenced by the Cr content? If so, how? 2. Does Cr change the stability of radiation-produced defects? 3. Is the diffusivity of cascade-produced defects changed by Cr content? 4. How do Cr atoms redistribute under irradiation inside the material under the action of thermodynamic driving forces and radiation-defect fluxes? It is found that the presence of Cr does not influence the type of damage created by displacement cascades, as compared to pure Fe, while cascades do contribute to redistributing Cr, in the same direction as thermodynamic driving forces. The presence of Cr does change the stability of point-defects: the effect is weak in the case of vacancies, stronger in the case of self-interstitials. In the latter case, Cr increases the stability of self-interstitial clusters, especially those so small to be invisible to the electron microscope. Cr reduces also significantly the diffusivity of self-interstitials and their clusters, in a way that depends in a non-monotonic way on Cr content, as well as on cluster size and temperature; however, the effect is negligible on the mobility of self-interstitial clusters large enough to become visible dislocation loops. Finally, Cr-rich precipitate formation is favoured in the tensile region of edge dislocations, while it appears not to be influenced by screw dislocations; prismatic dislocation loops

  20. Phase stability under irradiation in alloys with a positive heat of mixing: Effective thermodynamics description

    SciTech Connect

    Enrique, Raul A.; Bellon, Pascal

    1999-12-01

    The alteration of phase stability due to the continuous production of forced atomic displacements, as is the case under irradiation, is studied. A simple kinetic model of a binary alloy exhibiting phase separation is investigated, and two limiting cases are considered: nearest-neighbor ballistic exchanges and arbitrary-length ballistic exchanges. The model is simultaneously studied by direct kinetic Monte Carlo simulations, and by theoretical approaches based in the description of the steady state by effective thermodynamics. Two theoretical frameworks are considered: a microscopic description based in effective interactions, and a macroscopic description based in effective free energies constructed over a modified Cahn-Hilliard equation. Developments of these models are proposed to allow for quantitative predictions. For nearest-neighbor ballistic exchanges, the steady-state phase diagram is evaluated for each of the approaches and the results are directly compared. In the case of arbitrary-length ballistic exchanges, the appearance of labyrinthine mesoscopic structures at steady state is rationalized in terms of the competition between short-range attractive and long-range repulsive effective interactions. The favorable comparison between the theoretical results and the direct Kinetic Monte Carlo simulations shows that the long-term behavior of this inherently far-from-equilibrium system can be described in terms of effective thermodynamic potentials. (c) 1999 The American Physical Society.

  1. Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

    DOE PAGES

    Jin, Ke; Guo, Wei; Lu, Chenyang; Ullah, Mohammad W.; Zhang, Yanwen; Weber, William J.; Wang, Lumin; Poplawsky, Jonathan D.; Bei, Hongbin

    2016-12-01

    Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 1013 to 3 × 1016 cm-2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluence regime, which ismore » consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

  2. Effect of Cr content on the nanostructural evolution of irradiated ferritic/martensitic alloys: An object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Malerba, L.; Becquart, C. S.

    2015-10-01

    Self-interstitial cluster diffusivity in Fe-Cr alloys, model materials for high-Cr ferritic/martensitic steels, is known to be reduced in a non-monotonic way as a function of Cr concentration: it first decreases, then increases. This non-monotonic behaviour is caused by a relatively long-ranged attractive interaction between Cr atoms and crowdions and correlates well with the experimentally observed swelling in these alloys under neutron irradiation, also seen to first decrease and then increase with increasing Cr content, under comparable irradiation conditions. Moreover, recent studies reveal that C atoms dispersed in the Fe matrix form under irradiation complexes with vacancies which, in turn, act as trap for one-dimensionally migrating self-interstitial clusters. The mobility of one-dimensional migrating clusters is considered key to determine swelling susceptibility. However, no model has ever been built that quantitatively describes the dependence of swelling on Cr content, allowing for the presence of C in the matrix. In this work we developed physically-based sets of parameters for object kinetic Monte Carlo (OKMC) simulations intended to study the nanostructure evolution under irradiation in Fe-Cr-C alloys. The nanostructural evolution in Fe-C and in four Fe-Cr-C alloys (containing 2.5, 5, 9 and 12 wt.% Cr) neutron irradiated up to ∼0.6 dpa at 563 K was simulated according to the model and reference experiments were reproduced. Our model shows that the SIA cluster reduced mobility has a major influence on the nanostructural evolution: it increases the number of vacancy-SIA recombinations and thus leads to the suppression of voids formation. This provides a clear framework to interpret the non-monotonic dependence of swelling in Fe-Cr alloys versus Cr content. Our model also suggests that the amount of C in the matrix has an equally important role: high amounts of it may counteract the beneficial effect that Cr has in reducing swelling.

  3. The effect of prolonged irradiation on defect production and ordering in Fe-Cr and Fe-Ni alloys.

    PubMed

    Vörtler, K; Juslin, N; Bonny, G; Malerba, L; Nordlund, K

    2011-09-01

    The understanding of the primary radiation damage in Fe-based alloys is of interest for the use of advanced steels in future fusion and fission reactors. In this work Fe-Cr alloys (with 5, 6.25, 10 and 15% Cr content) and Fe-Ni alloys (with 10, 40, 50 and 75% Ni content) were used as model materials for studying the features of steels from a radiation damage perspective. The effect of prolonged irradiation (neglecting diffusion), i.e. the overlapping of single 5 keV displacement cascade events, was studied by molecular dynamics simulation. Up to 200 single cascades were simulated, randomly induced in sequence in one simulation cell, to study the difference between fcc and bcc lattices, as well as initially ordered and random crystals. With increasing numbers of cascades we observed a saturation of Frenkel pairs in the bcc alloys. In fcc Fe-Ni, in contrast, we saw a continuous accumulation of defects: the growth of stacking-fault tetrahedra and a larger number of self-interstitial atom clusters were seen in contrast to bcc alloys. For all simulations the defect clusters and the short range order parameter were analysed in detail depending on the number of cascades in the crystal. We also report the modification of the repulsive part of the Fe-Ni interaction potential, which was needed to study the non-equilibrium processes. PMID:21846941

  4. Creating poly(ethylene glycol) film on the surface of NiTi alloy by gamma irradiation

    NASA Astrophysics Data System (ADS)

    Yu, Hongyan; Yan, Jin; Ma, Huiling; Zeng, Xinmiao; Liu, Yang; Zhao, Xinqing

    2015-07-01

    NiTi alloy has been extensively utilized as biomaterials owing to its unique shape memory effect, superelasticity and biocompatibility. However, concern with the toxic and allergic responses of nickel potentially releasing from implants stimulated lots of researches of modification on NiTi alloy surface. Creating chemical bond attachment of bioorganic film on NiTi alloy surface could effectively inhibit Ni releasing and obtain bioactive functions for further application. In this work, to get a bioorganic surface, NiTi alloy was modified with poly(ethylene glycol) (PEG) film by gamma ray induced grafting or crosslinking. X-ray diffraction (XRD) spectrum, water contact angle geometer and X-ray photoelectron spectroscopy (XPS) techniques were used to characterize the NiTi surface. The results indicated that PEG was covalent bonded on NiTi alloy surface. Fluorescence microscope (FM) images for morphology of 1 day osteoblast culture on the PEG coated NiTi surface showed that PEG could improve cell proliferation on NiTi surface. Our work offers a way to introduce a bioorganic metal surface by gamma irradiation.

  5. The Kinetics of Dislocation Loop Formation in Ferritic Alloys Through the Aggregation of Irradiation Induced Defects

    NASA Astrophysics Data System (ADS)

    Kohnert, Aaron Anthony

    The mechanical properties of materials are often degraded over time by exposure to irradiation environments, a phenomenon that has hindered the development of multiple nuclear reactor design concepts. Such property changes are the result of microstructural changes induced by the collision of high energy particles with the atoms in a material. The lattice defects generated in these recoil events migrate and interact to form extended damage structures. This study has used theoretical models based on the mean field chemical reaction rate theory to analyze the aggregation of isolated lattice defects into larger microstructural features that are responsible for long term property changes, focusing on the development of black dot damage in ferritic iron based alloys. The purpose of such endeavors is two-fold. Primarily, such models explain and quantify the processes through which these microstructures form. Additionally, models provide insight into the behavior and properties of the point defects and defect clusters which drive general microstructural evolution processes. The modeling effort presented in this work has focused on physical fidelity, drawing from a variety of sources of information to characterize the unobservable defect generation and agglomeration processes that give rise to the observable features reported in experimental data. As such, the models are based not solely on isolated point defect creation, as is the case with many older rate theory approaches, but instead on realistic estimates of the defect cluster population produced in high energy cascade damage events. Experimental assessments of the microstructural changes evident in transmission electron microscopy studies provide a means to measure the efficacy of the kinetic models. Using common assumptions of the mobility of defect clusters generated in cascade damage conditions, an unphysically high density of damage features develops at the temperatures of interest with a temperature dependence

  6. The interaction of point defects with line dislocations in HVEM (high voltage electron microscope) irradiated Fe-Ni-Cr alloys

    SciTech Connect

    King, S.L.; Jenkins, M.L. . Dept. of Materials); Kirk, M.A. ); English, C.A. . Materials Development Div.)

    1990-05-01

    This paper presents results of a study of the interaction of point defects produced by high voltage electron microscope (HVEM) irradiation with pre-existing dislocations in austenitic Fe-15% 25%Ni-17%Cr alloys, aimed at the determination of the mechanisms of climb of dissociated dislocations. Dislocations were initially characterized at sub-threshold voltages (here 200kV) using the weak-beam technique. These dislocations were then irradiated with 1MeV electrons in the Argonne HVEM before being returned to a lower voltage microscope for post-irradiation characterization. Interstitial climb was seen only at particularly favorable sites, such as pre-existing jogs, whilst vacancies clustered near dislocations, forming stacking fault tetrahedra (SFT). Partial separations were also observed to have decreased after irradiation. The post-irradiation configuration was found to depend strongly on both dislocation character and pre-irradiation dislocation configuration. These results, and their relevance to the void swelling problem, are discussed. 52 refs., 8 figs.

  7. Evaluation of ferritic alloy Fe-2-1/4Cr-1Mo after neutron irradiation - irradiation creep and swelling

    SciTech Connect

    Gelles, D.S.; Puigh, R.J.

    1983-10-01

    Irradiation creep and swelling measurements are reported for Fe-2-1/4Cr-1Mo after irradiation by fast neutrons over the temperature range 390 to 560/sup 0/C. Diameter change measurements on thin walled pressurized tubes in a bainitic condition and density change measurements on rods in a nonstandard condition were made following irradiation in the Experimental Breeder Reactor II. The irradiation creep specimens were irradiated to a fluence of 5.7 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) or 30 dpa and the swelling specimens were irradiated to a peak fluence of 2.4 x 10/sup 23/ n/cm/sup 2/ or 115 dpa. These results have been used as a basis to establish in-reactor creep and swelling correlations for 2-1/4Cr-1Mo in a bainitic condition. The correlations predict moderate swelling and moderate irradiation enhanced creep at 390/sup 0/C.

  8. Understanding of copper precipitation under electron or ion irradiations in FeCu0.1 wt% ferritic alloy by combination of experiments and modelling

    NASA Astrophysics Data System (ADS)

    Radiguet, B.; Barbu, A.; Pareige, P.

    2007-02-01

    This work is dedicated to the understanding of the basic processes involved in the formation of copper enriched clusters in low alloyed FeCu binary system (FeCu0.1 wt%) under irradiation at temperature close to 300 °C. Such an alloy was irradiated with electrons or with ions (Fe+ or He+) in order to deconvolute the effect of displacement cascades and the associated generation of point defect clusters (ion irradiations), and the super-saturation of mono-vacancies and self-interstitial atoms (electron irradiation). The microstructure of this alloy was characterised by tomographic atom probe. Experimental results were compared with results obtained with cluster dynamic model giving an estimation of the evolution of point defects (free or agglomerated) under irradiation on the one hand and describing homogeneous enhanced precipitation of copper on the other hand. The comparison between the results obtained on the different irradiation conditions and the model suggests that the point defect clusters (dislocation loops and/or nano-voids) created in displacement cascades play a major role in copper clustering in low copper alloy irradiated at 573 K.

  9. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    NASA Astrophysics Data System (ADS)

    Chernov, I. I.; Kalashnikov, A. N.; Kalin, B. A.; Binyukova, S. Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 × 10 20 m -2 at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content ( NC>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  10. Mechanism of instability of carbides in Fe-TaC alloy under high energy electron irradiation at 673 K

    NASA Astrophysics Data System (ADS)

    Abe, Hiroaki; Ishizaki, Takahiro; Kano, Sho; Li, Feng; Satoh, Yuhki; Tanigawa, Hiroyasu; Hamaguchi, Dai; Nagase, Takeshi; Yasuda, Hidehiro

    2014-12-01

    Reduced activation ferritic/martensitic steels (RAFMs), such as F82H steel, are designed to enhance the high-temperature strength by formation of MX-type nanometer-scale precipitates, mainly TaC. However, their instability under irradiation was recently reported. The purpose of this work, therefore, is to clarify the mechanism employing simultaneous observations under electron irradiation at elevated temperature in a high voltage electron microscope. In this work, Fe-0.2 wt.% TaC was fabricated as a model alloy of F82H steel. The instability of the precipitates was observed under electron irradiation at 1 MeV or above. The remarkable shrinkage and disappearance were clearly observed under irradiation with 1.5 MeV and above. On the contrary, the precipitates were mostly stable below 0.75 MeV. Two kinds of mechanism of the irradiation-induced instability were deduced from the electron-energy dependence. One is the dissolution and diffusion of tantalum from precipitates in ferrite matrix. The other is the displacements of tantalum in precipitates that introduce dissolution of Ta into matrix.

  11. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    DOE PAGES

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; et al

    2016-02-01

    We report that energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters farmore » exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.« less

  12. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys.

    PubMed

    Lu, Chenyang; Jin, Ke; Béland, Laurent K; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M; Stoller, Roger E; Wang, Lumin

    2016-01-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

  13. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys.

    PubMed

    Lu, Chenyang; Jin, Ke; Béland, Laurent K; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M; Stoller, Roger E; Wang, Lumin

    2016-01-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance. PMID:26829570

  14. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    NASA Astrophysics Data System (ADS)

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-02-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

  15. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    PubMed Central

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-01-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance. PMID:26829570

  16. Formation and dissolution of oxide film on zirconium alloys in 288°C pure water under γ-ray irradiation

    NASA Astrophysics Data System (ADS)

    Nishino, Yoshitaka; Endo, Masao; Ibe, Eishi; Yasuda, Takayoshi

    1997-09-01

    Laboratory corrosion tests for zirconium alloys which were based on Zircaloy-2 were performed in 288°C oxygenated pure water for 100 days both with and without 60Co γ-ray irradiation. No nodular oxide was observed. Corrosion weight gains of the alloy which had the lowest nodular corrosion resistance were lower for the irradiated condition than the non-irradiated condition. On the other hand, the alloys which had higher nodular corrosion resistances showed almost the same weight gains for both conditions. Differences of weight gain with and without irradiation were attributed to dissolution of the oxide film in the high temperature water. Dissolution tests of single crystal yttria-doped ZrO 2 indicated that 30-40% larger amounts dissolved into water under the γ-ray irradiation. Low angle incident X-ray diffraction analysis showed that the tetragonal-ZrO 2 fraction in the oxide film was lower with irradiation than without it, especially for the near surface area. The water radiolysis species accelerated the dissolution of the oxide film, especially film on the alloy with lower nodular corrosion resistance. This dissolution led to the lower tetragonal-ZrO 2 fraction and was considered to be one of the factors causing a localized breakdown of the barrier oxide film to make the nodular oxide.

  17. Formation and evolution of intermetallic nanoparticles and vacancy defects under irradiation in Fesbnd Nisbnd Al ageing alloy characterized by resistivity measurements and positron annihilation

    NASA Astrophysics Data System (ADS)

    Druzhkov, A. P.; Danilov, S. E.; Perminov, D. A.; Arbuzov, V. L.

    2016-08-01

    In this paper, we study the effects of intermetallic nanoparticles like Ni3Al on the evolution of vacancy defects in the fcc Fesbnd Nisbnd Al alloy under electron irradiation using positron annihilation spectroscopy. Electrical resistivity measurements have been used as a testing method for characterizing the evolution in the underlying precipitate microstructure due to heat treatment and irradiation. It was shown that the nanosized (∼4.5 nm) intermetallic precipitates homogeneously distributed in the alloy matrix caused a several-fold decrease in the accumulation of vacancies as compared to their accumulation in the pre-quenched alloy. This effect was enhanced with the irradiation temperature. The irradiation-induced growth of intermetallic nanoparticles was also observed in the pre-quenched Fesbnd Nisbnd Al alloy under irradiation at 573 K. Thus, resistivity measurement and positron confinement in ultrafine intermetallic particles, which we revealed earlier, provided the control over the evolution of coherent precipitates, along with vacancy defects, during irradiation and annealing.

  18. The microstructure of rapidly solidified path a prime candidate alloys following irradiation with Fe and He ions

    NASA Astrophysics Data System (ADS)

    Arnberg, L.; Vander Sande, J. B.; Frost, H. J.; Harling, O. K.

    A rapidly solidified,titanium modified, cold worked Type 316 stainless steel has been irradiated simultaneously with Fe and He ions at temperatures between 450 and 650°C. The material shows no void swelling at 450°C even at 50 dpa. At 550°C the swelling is very low below 40 dpa (<0.01%) but increases with fluence to 0.4% at 60 dpa. At 650°C both the cavity number density and size increase to give a swelling of 0.8% at 40 dpa. A conventionally ingot cast alloy with the same composition shows comparable swelling but the void number density in this material is significantly higher. Small (<200 Å) TiC particles which were originally present in the rapidly solidified material are dissolved at fluences above 40 dpa at all temperatures. A silicon and nickel rich phase has precipitated during irradiations at 550 and 650°C.

  19. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  20. Revised ANL-reported tensile data for unirradiated and irradiated (FFTF, HFIR) V-Ti and V-Cr-Ti alloys

    SciTech Connect

    Billone, M.C.

    1998-03-01

    The tensile data for all unirradiated and irradiated vanadium alloys samples tested at Argonne National Laboratory (ANL) have been critically reviewed and, when necessary, revised. The review and revision are based on reanalyzing the original load-displacement strip chart recordings by a methodology consistent with current ASTM standards. For unirradiated alloys (162 samples), the revised values differ from the previous values as follows: {minus}11{+-}19 MPa ({minus}4{+-}6%) for yield strength (YS), {minus}3{+-}15 MPa ({minus}1{+-}3%) for ultimate tensile strength (UTS), {minus}5{+-}2% strain for uniform elongation (UE), and {minus}4{+-}2% strain for total elongation (TE). Of these changes, the decrease in {minus}1{+-}6 MPa (0{+-}1%) for UTS, {minus}5{+-}2% for UE, and {minus}4{+-}2% for TE. Of these changes, the decrease in UE values for alloys irradiated and tested at 400--435 C is the most significant. This decrease results from the proper subtraction of nongauge-length deformation from measured crosshead deformation. In previous analysis of the tensile curves, the nongauge-length deformation was not correctly determined and subtracted from the crosshead displacement. The previously reported and revised tensile values for unirradiated alloys (20--700 C) are tabulated in Appendix A. The revised tensile values for the FFTF-irradiated (400--600 C) and HFIR-irradiated (400 C) alloys are tabulated in Appendix B, along with the neutron damage and helium levels. Appendix C compares the revised values to the previously reported values for irradiated alloys. Appendix D contains previous and revised values for the tensile properties of unirradiated V-5Cr-5Ti (BL-63) alloy exposed to oxygen.

  1. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    NASA Astrophysics Data System (ADS)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  2. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    SciTech Connect

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  3. Irradiation effect of nano-bubble dispersion strengthened (N-BDS) alloy

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Kawano, Ryohei; Shi, Shi; Ukai, Shigeharu; Hayashi, Shigenari; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2013-11-01

    Nano-bubble dispersion strengthened (N-BDS) Fe was made from Fe and polymethylmethacrylate (PMMA) powder and irradiated by 6.4 MeV Fe3+ ions to investigate the cavity strengthening and the bubble to void evolution. The bubbles accelerated the irradiation-induced cavity growth. The hardness of the N-BDS Fe was 500 MPa higher than that of unalloyed Fe and the hardness increased by irradiation, while that of unalloyed Fe did not increase. Cavity is probably the origin of the irradiation hardening of N-BDS Fe.

  4. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

    SciTech Connect

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

  5. Phase-field Model for Interstitial Loop Growth Kinetics and Thermodynamic and Kinetic Models of Irradiated Fe-Cr Alloys

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Sun, Xin; Khaleel, Mohammad A.

    2011-06-15

    strength of interstitial loop for interstitials. In part II, we present a generic phase field model and discuss the thermodynamic and kinetic properties in phase-field models including the reaction kinetics of radiation defects and local free energy of irradiated materials. In particular, a two-sublattice thermodynamic model is suggested to describe the local free energy of alloys with irradiated defects. Fe-Cr alloy is taken as an example to explain the required thermodynamic and kinetic properties for quantitative phase-field modeling. Finally the great challenges in phase-field modeling will be discussed.

  6. Defect evolution in a Nisbnd Mosbnd Crsbnd Fe alloy subjected to high-dose Kr ion irradiation at elevated temperature

    NASA Astrophysics Data System (ADS)

    de los Reyes, Massey; Voskoboinikov, Roman; Kirk, Marquis A.; Huang, Hefei; Lumpkin, Greg; Bhattacharyya, Dhriti

    2016-06-01

    A candidate Nisbnd Mosbnd Crsbnd Fe alloy (GH3535) for application as a structural material in a molten salt nuclear reactor was irradiated with 1 MeV Kr2+ ions (723 K, max dose of 100 dpa) at the IVEM-Tandem facility. The evolution of defects like dislocation loops and vacancy- and self-interstitial clusters was examined in-situ. For obtaining a deeper insight into the true nature of these defects, the irradiated sample was further analysed under a TEM post-facto. The results show that there is a range of different types of defects formed under irradiation. Interaction of radiation defects with each other and with pre-existing defects, e.g., linear dislocations, leads to the formation of complex microstructures. Molecular dynamics simulations used to obtain a greater understanding of these defect transformations showed that the interaction between linear dislocations and radiation induced dislocation loops could form faulted structures that explain the fringed contrast of these defects observed in TEM.

  7. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

    SciTech Connect

    Long, Fei; Daymond, Mark R. Yao, Zhongwen; Kirk, Marquis A.

    2015-03-14

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.

  8. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Long, Fei; Daymond, Mark R.; Yao, Zhongwen; Kirk, Marquis A.

    2015-03-01

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, {01 1 ¯ 1 }⟨0 1 ¯ 12 ⟩ twinning was identified in the irradiated sample deformed at 300 °C.

  9. Microstructural evolution in NF616 (P92) and Fe–9Cr–0.1C-model alloy under heavy ion irradiation

    SciTech Connect

    Topbasi, Cem; Kaoumi, Djamel; Motta, Arthur T.; Kirk, Mark A.

    2015-11-01

    In this comparative study, in situ investigations of the microstructure evolution in a Fee9Cr ferritic emartensitic steel, NF616, and a Fee9Cre0.1C-model alloy with a similar ferriticemartensitic microstructure have been performed. NF616 and Fee9Cre0.1C-model alloy were irradiated to high doses (up to ~10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fee9Cre0.1C-model alloy around ~1 dpa. At low temperatures (50e298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fee9Cre0.1C-model alloy around ~2e3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ~5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fee9Cre0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fee9Cre0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

  10. Microstructural evolution in NF616 (P92) and Fe-9Cr-0.1C-model alloy under heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Topbasi, Cem; Kaoumi, Djamel; Motta, Arthur T.; Kirk, Mark A.

    2015-11-01

    In this comparative study, in situ investigations of the microstructure evolution in a Fe-9Cr ferritic-martensitic steel, NF616, and a Fe-9Cr-0.1C-model alloy with a similar ferritic-martensitic microstructure have been performed. NF616 and Fe-9Cr-0.1C-model alloy were irradiated to high doses (up to ∼10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fe-9Cr-0.1C-model alloy around ∼1 dpa. At low temperatures (50-298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fe-9Cr-0.1C-model alloy around ∼2-3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ∼5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fe-9Cr-0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fe-9Cr-0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

  11. Enhancing the absorption of aluminum alloys by irradiation with an excimer laser

    NASA Astrophysics Data System (ADS)

    Scott, Graeme; Williams, Stewart W.; Morgan, P. C.; Dempster, M.

    1994-09-01

    Aluminum alloys typically have as received reflectivities of 85 - 95% at 10.6 micrometers making many laser processes difficult or impossible. These values have been reduced to as low as 1 - 2% by optimizing the processing parameters of an excimer laser used to modify the surface structure of 8090 and 2024 Al alloys and pure Al prior to their exposure to a CO2 laser. The most significant excimer processing parameters were found to be the scan pattern of the excimer beam, the number of pulses per scan pattern step (dwell time) and the laser fluence. Optimizing these parameters allows the production of a rough oxide rich surface and reflectivities at 10.6 micrometers routinely below 10%. Preliminary results are presented from the practical implementation of the technique to a dual wavelength (CO2/excimer) cutting system. Increases in cutting speeds of between 2 - 4 times are demonstrated with 8090 Al-Li alloy using dual wavelength laser processing.

  12. Investigation of the thermo-mechanical behavior of neutron-irradiated Fe-Cr alloys by self-consistent plasticity theory

    NASA Astrophysics Data System (ADS)

    Xiao, Xiazi; Terentyev, Dmitry; Yu, Long; Bakaev, A.; Jin, Zhaohui; Duan, Huiling

    2016-08-01

    The thermo-mechanical behavior of non-irradiated (at 223 K, 302 K and 573 K) and neutron irradiated (at 573 K) Fe-2.5Cr, Fe-5Cr and Fe-9Cr alloys is studied by a self-consistent plasticity theory, which consists of constitutive equations describing the contribution of radiation defects at grain level, and the elastic-viscoplastic self-consistent method to obtain polycrystalline behaviors. Attention is paid to two types of radiation-induced defects: interstitial dislocation loops and solute rich clusters, which are believed to be the main sources of hardening in Fe-Cr alloys at medium irradiation doses. Both the hardening mechanism and microstructural evolution are investigated by using available experimental data on microstructures, and implementing hardening rules derived from atomistic data. Good agreement with experimental data is achieved for both the yield stress and strain hardening of non-irradiated and irradiated Fe-Cr alloys by treating dislocation loops as strong thermally activated obstacles and solute rich clusters as weak shearable ones.

  13. In situ observation of defect annihilation in Kr ion-irradiated bulk Fe/amorphous-Fe 2 Zr nanocomposite alloy

    SciTech Connect

    Yu, K. Y.; Fan, Z.; Chen, Y.; Song, M.; Liu, Y.; Wang, H.; Kirk, M. A.; Li, M.; Zhang, X.

    2014-08-26

    Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe96Zr4 nanocomposite alloy. Irradiation resulted in amorphization of Fe2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphous nanocomposites.

  14. Investigation of the radiation resistance of triple-junction a-Si:H alloy solar cells irradiated with 1.00 MeV protons

    NASA Technical Reports Server (NTRS)

    Lord, Kenneth R., II; Walters, Michael R.; Woodyard, James R.

    1993-01-01

    The effect of 1.00 MeV proton irradiation on hydrogenated amorphous silicon alloy triple-junction solar cells is reported for the first time. The cells were designed for radiation resistance studies and included 0.35 cm(sup 2) active areas on 1.0 by 2.0 cm(sup 2) glass superstrates. Three cells were irradiated through the bottom contact at each of six fluences between 5.10E12 and 1.46E15 cm(sup -2). The effect of the irradiations was determined with light current-voltage measurements. Proton irradiation degraded the cell power densities from 8.0 to 98 percent for the fluences investigated. Annealing irradiated cells at 200 C for two hours restored the power densities to better than 90 percent. The cells exhibited radiation resistances which are superior to cells reported in the literature for fluences less than 1E14 cm(sup -2).

  15. Correlation of radiation-induced changes in mechanical properties and microstructural development of Alloy 718 irradiated with mixed spectra of high-energy protons and spallation neutrons

    NASA Astrophysics Data System (ADS)

    Sencer, B. H.; Bond, G. M.; Garner, F. A.; Hamilton, M. L.; Maloy, S. A.; Sommer, W. F.

    2001-07-01

    Alloy 718 is a γ '(Ni 3(Al,Ti))-γ″(Ni 3Nb) hardenable superalloy with attractive strength, and corrosion resistance. This alloy is a candidate material for use in accelerator production of tritium (APT) target and blanket applications, where it would have to withstand low-temperature irradiation by high-energy protons and spallation neutrons. The existing data base, relevant to such irradiation conditions, is very limited. Alloy 718 has therefore been exposed to a particle flux and spectrum at the Los Alamos Neutron Science Center (LANSCE), closely matching those expected in the APT target and blanket applications. The yield stress of Alloy 718 increases with increasing dose up to ˜0.5 dpa, and then decreases with further increase in dose. The uniform elongation, however, drastically decreases with increasing dose at very low doses (<0.5 dpa), and does not recover when the alloy later softens somewhat. Transmission electron microscopy (TEM) investigation of Alloy 718 shows that superlattice spots corresponding to the age-hardening precipitate phases γ ' and γ″ are lost from the diffraction patterns for Alloy 718 by only 0.6 dpa, the lowest proton-induced dose level achieved in this experiment. Examination of samples that were neutron irradiated to doses of only ˜0.1 dpa showed that precipitates are faintly visible in diffraction patterns but are rapidly becoming invisible. It is proposed that the γ ' and γ″ first become disordered (by <0.6 dpa), but remain as solute-rich aggregates that still contribute to the hardness at relatively low dpa levels, and then are gradually dispersed at higher doses.

  16. The independence of irradiation creep in austenitic alloys of displacement rate and helium to dpa ratio

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.; Grossbeck, M.L.

    1997-04-01

    The majority of high fluence data on the void swelling and irradiation creep of austenitic steels were generated at relatively high displacement rates and relatively low helium/dpa levels that are not characteristic of the conditions anticipated in ITER and other anticipated fusion environments. After reanalyzing the available data, this paper shows that irradiation creep is not directly sensitive to either the helium/dpa ratio or the displacement rate, other than through their possible influence on void swelling, since one component of the irradiation creep rate varies with no correlation to the instantaneous swelling rate. Until recently, however, the non-swelling-related creep component was also thought to exhibit its own strong dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale at temperatures in the 270-350{degrees}C range. More recently this perception was thought to extend to higher irradiation temperatures. It now appears, however, that this interpretation is incorrect, and in fact the steady-state value of the non-swelling component of irradiation creep is actually insensitive to displacement rate. The perceived flux dependence appears to arise from a failure to properly interpret the impact of the transient regime of irradiation creep.

  17. Weak-beam imaging of dissociated dislocations in HVEM-irradiated Fe-Ni-Cr alloys

    SciTech Connect

    King, S.L.; Jenkins, M.L.; Kirk, M.A.; English, C.A.

    1992-06-01

    We report here on studies by weak-beam electron microscopy of the evolution of microstructures at and near preexisting line dislocations in a number of Fe-Ni-Cr alloys under electronirradiation in a high-voltage electron microscope (HVEM). The detailed observations are discussed in terms of dislocation climb mechanisms in these materials and a model based on interstitial pipe diffusion.

  18. Alloy

    NASA Astrophysics Data System (ADS)

    Cabeza, Sandra; Garcés, Gerardo; Pérez, Pablo; Adeva, Paloma

    2014-07-01

    The Mg98.5Gd1Zn0.5 alloy produced by a powder metallurgy route was studied and compared with the same alloy produced by extrusion of ingots. Atomized powders were cold compacted and extruded at 623 K and 673 K (350 °C and 400 °C). The microstructure of extruded materials was characterized by α-Mg grains, and Mg3Gd and 14H-LPSO particles located at grain boundaries. Grain size decreased from 6.8 μm in the extruded ingot, down to 1.6 μm for powders extruded at 623 K (350 °C). Grain refinement resulted in an increase in mechanical properties at room and high temperatures. Moreover, at high temperatures the PM alloy showed superplasticity at high strain rates, with elongations to failure up to 700 pct.

  19. Microstructure evolution and hardness change in ordered Ni3V intermetallic alloy by energetic ion irradiation

    NASA Astrophysics Data System (ADS)

    Hashimoto, A.; Kaneno, Y.; Semboshi, S.; Yoshizaki, H.; Saitoh, Y.; Okamoto, Y.; Iwase, A.

    2014-11-01

    Ni3V bulk intermetallic compounds with ordered D022 structure were irradiated with 16 MeV Au ions at room temperature. The irradiation induced phase transformation was examined by means of the transmission electron microscope (TEM), the extended X-ray absorption fine structure measurement (EXAFS) and the X-ray diffraction (XRD). We also measured the Vickers hardness for unirradiated and irradiated specimens. The TEM observation shows that by the Au irradiation, the lamellar microstructures and the super lattice spot in diffraction pattern for the unirradiated specimen disappeared. This TEM result as well as the result of XRD and EXAFS measurements means that the intrinsic D022 structure of Ni3V changes into the A1 (fcc) structure which is the lattice structure just below the melting point in the thermal equilibrium phase diagram. The lattice structure change from D022 to A1 (fcc) accompanies a remarkable decrease in Vickers microhardness. The change in crystal structure was discussed in terms of the thermal spike and the sequential atomic displacements induced by the energetic heavy ion irradiation.

  20. Effects of proton irradiation on nanocluster precipitation in ferritic steel containing fcc alloying additions

    SciTech Connect

    Zhang, Zhongwu; Liu, C T; Wang, Xun-Li; Miller, Michael K; Ma, Dong; Chen, Guang; Williams, J R; Chin, Bryan

    2012-01-01

    Newly-developed precipitate-strengthened ferritic steels with and without pre-existing nanoscale precipitates were irradiated with 4 MeV protons to a dose of ~5 mdpa at 50 C and subsequently examined by nanoindentation and atom probe tomography (APT). Irradiation-enhanced precipitation and coarsening of pre-existing nanoscale precipitates were observed. Copper partitions to the precipitate core along with a segregation of Ni, Al and Mn to the precipitate/matrix interface after both thermal aging and proton irradiation. Proton irradiation induces the precipitation reaction and coarsening of pre-existing nanoscale precipitates, and these results are similar to a thermal aging process. The precipitation and coarsening of nanoscale precipitates are responsible for the changes in hardness. The observation of the radiation-induced softening is essentially due to the coarsening of the pre-existing Cu-rich nanoscale precipitates. The implication of the precipitation on the embrittlement of reactor-pressure-vessel steels after irradiation is discussed.

  1. Nucleation of Cr precipitates in Fe-Cr alloy under irradiation

    SciTech Connect

    Dai, Y. Y.; Ao, L.; Sun, Qing- Qiang; Yang, L.; Nie, JL; Peng, SM; Long, XG; Zhou, X. S.; Zu, Xiaotao; Liu, L.; Sun, Xin; Terentyev, Dimtry; Gao, Fei

    2015-04-01

    The nucleation of Cr precipitates induced by overlapping of displacement cascades in Fe-Cr alloys has been investigated using the combination of molecular dynamics (MD) and Metropolis Monte Carlo (MMC) simulations. The results reveal that the number of Frenkel pairs increases with the increasing of overlapped cascades. Overlapping cascades could promote the formation of Cr precipitates in Fe-Cr alloys, as analyzed using short range order (SRO) parameters to quantify the degree of ordering and clustering of Cr atoms. In addition, the simulations using MMC approach show that the presence of small Cr clusters and vacancy clusters formed within cascade overlapped region enhance the nucleation of Cr precipitates, leading to the formation of large Cr dilute precipitates.

  2. Shear punch testing of {sup 59}Ni isotopically-doped model austenitic alloys after irradiation in FFTF at different He/dpa ratios

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Hamilton, M.L.; Garner, F.A.

    1998-03-01

    A series of three model alloys, Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr45Ni were irradiated side-by-side in FFTF-MOTA in both the annealed and the cold worked condition in each of two variants, one using naturally occurring isotopic mixtures, and another doped with {sup 59}Ni to generate relatively high helium-to-dpa ratios. Previous papers in this series have addressed the influence of helium on radiation-induced evolution of microstructure, dimensional stability and mechanical properties, the latter using miniature-tensile specimens. In the final paper of this experimental series, three sets of irradiations conducted at different temperatures and displacement rates were examined by shear punch testing of standard microscopy disks. The results were used to determine the influence of helium generation rate, alloy starting condition, irradiation temperature and total neutron exposure. The results were also compared with the miniature tensile data obtained earlier. In general, all alloys approached saturation levels of strength and ductility that were relatively independent of He/dpa ratio and starting condition, but were sensitive to the irradiation temperature and total exposure. Some small influence of helium/dpa ratio on the shear strength is visible in the two series that ran at {approximately}490 C, but is not evident at 365 C.

  3. Subtask 12H1: Vanadium alloy irradiation experiment X530 in EBR-II

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.; Chung, H.M.; Nowicki, L.J.; Smith, D.L.

    1995-03-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994. To obtain early irradiation performance data on the new 500-kg production heat of the V-4Cr-4Ti material before the scheduled EBR-II shutdown, an experiment, X530, was expeditiously designed and assembled. Charpy, compact tension, tensile and TEM specimens with different thermal mechanical treatments (TMTs), were enclosed in two capsules and irradiated in the last run of EBR-II, Run 170, from August 9 through September 27. For comparison, specimens from some of the previous heats were also included in the test. The accrued exposure was 35 effective full power days, yielding a peak damage of {approx}4 dpa in the specimens. The irradiation is now complete and the vehicle is awaiting to be discharged from EBR-II for postirradiation disassembly. 4 figs., 2 tabs.

  4. Vanadium alloy irradiation experiment X530 in EBR-II{sup *}

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.

    1995-04-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994.

  5. Transmission electron microscopy investigation of the microstructure of Fe-Cr alloys induced by neutron and ion irradiation at 300 °C

    NASA Astrophysics Data System (ADS)

    Hernández-Mayoral, M.; Heintze, C.; Oñorbe, E.

    2016-06-01

    Four Fe-Cr binary alloys, with Cr content from 2.5 up to 12wt%, were neutron or ion irradiated up to a dose of 0.6 dpa at 300 °C. The microstructural response to irradiation has been characterised using Transmission Electron Microscopy (TEM). Both, neutrons and ions, gave rise to the formation of dislocation loops. The most striking difference between ion and neutron irradiation is the distribution of these loops in the sample. Except for the lowest Cr content, loops are distributed mainly along grain boundaries and dislocations in the neutron irradiated samples. The inhomogeneous distribution of dislocation loops could be related to the presence of α‧ precipitates in the matrix. In contrast, a homogeneous distribution is observed in all ion irradiated samples. This important difference is attributed to the orders of magnitude difference in dose rate between these two irradiation conditions. Moreover, the density of loops depends non-monotonically on Cr content in case of neutron irradiation, while it seems to increase with Cr content for ion implantation. Differences are also observed in terms of cluster size, with larger sizes for neutron irradiation than for ion implantation, again pointing towards an effect of the dose rate.

  6. Local electronic effects and irradiation resistance in high-entropy alloys

    SciTech Connect

    Egami, Takeshi; Stocks, George Malcolm; Nicholson, Don; Khorgolkhuu, Od; Ojha, Madhusudan

    2015-01-01

    High-entropy alloys are multicomponent solid solutions in which various elements with different chemistries and sizes occupy the same crystallographic lattice sites. Thus, none of the atoms perfectly fit the lattice site, giving rise to considerable local lattice distortions and atomic-level stresses. These characteristics can be beneficial for performance under both radiation and in a high-temperature environment, making them attractive candidates as nuclear materials. We discuss electronic origin of the atomic-level stresses based upon first-principles calculations using a density functional theory approach.

  7. The dependence of irradiation creep in austenitic alloys on displacement rate and helium to dpa ratio

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.; Grossbeck, M.L.

    1998-03-01

    Before the parametric dependencies of irradiation creep can be confidently determined, analysis of creep data requires that the various creep and non-creep strains be separated, as well as separating the transient, steady-state, and swelling-driven components of creep. When such separation is attained, it appears that the steady-state creep compliance, B{sub o}, is not a function of displacement rate, as has been previously assumed. It also appears that the formation and growth of helium bubbles under high helium generation conditions can lead to a significant enhancement of the irradiation creep coefficient. This is a transient influence that disappears as void swelling begins to dominate the total strain, but this transient can increase the apparent creep compliance by 100--200% at relatively low ({le}20) dpa levels.

  8. Effects of composition and helium injection on dislocation loop development in pure FeNiCr alloys under Ni ion irradiation

    NASA Astrophysics Data System (ADS)

    Kimoto, Takayoshi

    1993-08-01

    Pure Fe35Ni7Cr, Fe45Ni7Cr, Fe40Ni13Cr and Fe45Ni15Cr alloys were irradiated by 4MeV Ni 2+ ions at 948 K to doses of about 0.05, 0.3 and 1.0 dpa without helium injection or with simultaneous helium injection. With increasing Ni content and decreasing Cr content, the diameter of radiation-induced dislocation loops increased, and the dose at which the dislocation loops disappeared to develop into dislocation networks decreased. The diameter of dislocation loops induced by Ni 2+ ions irradiation with simultaneous helium injection was larger than that without helium injection for the Fe35Ni7Cr and Fe45Ni7Cr alloys.

  9. Neutron-Induced Microstructural Evolution of Fe-15Cr-16Ni Alloys at ~400 C During Neutron Irradiation in the FFTF Fast Reactor

    SciTech Connect

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.; Wolfer, W. G.; Isobe, Yoshihiro

    2001-06-30

    An experiment conducted at ~400 degrees C on simple model austenitic alloys (Fe-15Cr-16Ni and Fe-15Cr-16Ni-0.25Ti, both with and without 500 appm boron) irradiated in the FFTF fast reactor at seven different dpa rates clearly shows that lowering of the atomic displacement rate leads to a pronounced reduction in the transient regime of void swelling. While the steady state swelling rate (~1%/dpa) of these alloys is unaffected by changes in the dpa rate, the transient regime of swelling can vary from <1 to ~60 dpa when the dpa rate varies over more than two orders of magnitude. This range of dpa rates covers the full span of fusion, PWR and fast reactor rates. The origin of the flux sensitivity of swelling arises first in the evolution of the Frank dislocation loop population, its unfaulting, and the subsequent evolution of the dislocation network. There also appears to be some flux sensitivity to the void nucleation process. Most interestingly, the addition of titanium suppresses the void nucleation process somewhat, but does not alter the duration of the transient regime of swelling or its sensitivity to dpa rate. Side-by-side irradiation of boron-modified model alloys in this same experiment shows that higher helium generation rates homogenize the swelling somewhat, but do not significantly change its magnitude or flux sensitivity. The results of this study support the prediction that austenitic alloys irradiated at PWR-relevant displacement rates will most likely swell more than when irradiated at higher rates characteristic of fast reactors. Thus, the use of swelling data accumulated in fast reactors may possibly lead to an under-prediction of swelling in lower-flux PWRs and fusion devices.

  10. Changes in cluster magnetism and suppression of local superconductivity in amorphous FeCrB alloy irradiated by Ar+ ions

    NASA Astrophysics Data System (ADS)

    Okunev, V. D.; Samoilenko, Z. A.; Szymczak, H.; Szewczyk, A.; Szymczak, R.; Lewandowski, S. J.; Aleshkevych, P.; Malinowski, A.; Gierłowski, P.; Więckowski, J.; Wolny-Marszałek, M.; Jeżabek, M.; Varyukhin, V. N.; Antoshina, I. A.

    2016-02-01

    We show that cluster magnetism in ferromagnetic amorphous Fe67Cr18B15 alloy is related to the presence of large, D=150-250 Å, α-(Fe Cr) clusters responsible for basic changes in cluster magnetism, small, D=30-100 Å, α-(Fe, Cr) and Fe3B clusters and subcluster atomic α-(Fe, Cr, B) groupings, D=10-20 Å, in disordered intercluster medium. For initial sample and irradiated one (Φ=1.5×1018 ions/cm2) superconductivity exists in the cluster shells of metallic α-(Fe, Cr) phase where ferromagnetism of iron is counterbalanced by antiferromagnetism of chromium. At Φ=3×1018 ions/cm2, the internal stresses intensify and the process of iron and chromium phase separation, favorable for mesoscopic superconductivity, changes for inverse one promoting more homogeneous distribution of iron and chromium in the clusters as well as gigantic (twice as much) increase in density of the samples. As a result, in the cluster shells ferromagnetism is restored leading to the increase in magnetization of the sample and suppression of local superconductivity. For initial samples, the temperature dependence of resistivity ρ(T)~T2 is determined by the electron scattering on quantum defects. In strongly inhomogeneous samples, after irradiation by fluence Φ=1.5×1018 ions/cm2, the transition to a dependence ρ(T)~T1/2 is caused by the effects of weak localization. In more homogeneous samples, at Φ=3×1018 ions/cm2, a return to the dependence ρ(T)~T2 is observed.

  11. Interface alloying due to Kr-irradiation on Ni/Si system

    NASA Astrophysics Data System (ADS)

    Datta, D.; Bhattacharyya, S. R.

    2007-03-01

    This study deals with an investigation of silicide formation at the interface of Ni thin film and Si substrate induced by 300 keV 84Kr2+ ions at a number of doses under ambient condition. The implanted samples were subsequently annealed using a rapid thermal annealing system (RTA) in vacuum environment. The modification of the interface and surface properties has been analyzed using various methods like scanning electron microscope with energy dispersive X-ray analysis (SEM/EDAX), X-ray diffraction (XRD), etc. The energy dispersive X-ray spectra (EDS) show a decrease of Ni content in the samples with increasing dose, which indicates the sputter erosion due to ion irradiation. The SEM micrographs show no considerable surface modification under ion bombardment but the XRD spectra clearly reveal the formation of different phases of Ni2Si and NiSi at lower doses which transforms to NiSi as dose increases.

  12. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, Jr., A. Burtron; Levy, Ira S.; Trimble, Dennis J.; Lanning, Donald D.; Gerber, Franna S.

    1990-01-01

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

  13. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, A.B. Jr.; Levy, I.S.; Trimble, D.J.; Lanning, D.D.; Gerber, F.S.

    1990-04-10

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-based materials is disclosed. Samples of zirconium-based materials having different compositions and/or fabrication methods are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280 to 316 C). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hydriding for the same materials when subject to in-reactor (irradiated) corrosion. 1 figure.

  14. Effects of irradiation temperature and dose rate on the mechanical properties of self-ion implanted Fe and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    Hardie, Christopher D.; Williams, Ceri A.; Xu, Shuo; Roberts, Steve G.

    2013-08-01

    Pure Fe and model Fe-Cr alloys containing 5, 10 and 14%Cr were irradiated with Fe+ ions at a maximum energy of 2 MeV to the same dose of 0.6 dpa at temperatures of 300 °C, 400 °C and 500 °C, and at dose rates corresponding to 6 × 10-4 dpa/s and 3 × 10-5 dpa/s. All materials exhibited an increase in hardness after irradiation at 300 °C. After irradiation at 400 °C, hardening was observed only in Fe-Cr alloys, and not in the pure Fe. After irradiation at 500 °C, no hardening was observed in any of the materials tested. For irradiations at both 300 °C and 400 °C, greater hardening was found in the Fe-Cr alloys irradiated at the lower dose rate. Transmission electron microscopy and atom probe tomography of Fe 5%Cr identified larger dislocation loop densities and sizes in the alloy irradiated with the high dose rate and Cr precipitation in the alloy irradiated with the low dose rate. Loss of defects at extended sinks such as dislocations and grain boundaries. Growth or shrinkage of defect clusters by the capture of point defects. Mutual annihilation by the recombination of a vacancy and interstitial. At low dose rates and/or high irradiation temperatures, reaction path (i) (sinks) dominates and at a high dose rates and/or low irradiation temperature reaction path (iii) (recombination) dominates [2]. The evolution of radiation damage such as dislocation loops and voids and phenomena such as radiation induced segregation, swelling and creep, depend on the fraction of point defects which migrate to sinks, recombine or cluster within the lattice and will be influenced by the reaction path that dominates the microstructural evolution of the material under irradiation.The relative proportions of these reaction types are directly dependent on the density and mobility of the defects, and hence dependent on dose rate and temperature. In iron, vacancy type defects are generally found to have significantly higher activation energy for migration compared to interstitial

  15. Effects of irradiation temperature and dose rate on the mechanical properties of self-ion implanted Fe and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    Hardie, Christopher D.; Williams, Ceri A.; Xu, Shuo; Roberts, Steve G.

    2013-08-01

    Pure Fe and model Fe-Cr alloys containing 5, 10 and 14%Cr were irradiated with Fe+ ions at a maximum energy of 2 MeV to the same dose of 0.6 dpa at temperatures of 300 °C, 400 °C and 500 °C, and at dose rates corresponding to 6 × 10-4 dpa/s and 3 × 10-5 dpa/s. All materials exhibited an increase in hardness after irradiation at 300 °C. After irradiation at 400 °C, hardening was observed only in Fe-Cr alloys, and not in the pure Fe. After irradiation at 500 °C, no hardening was observed in any of the materials tested. For irradiations at both 300 °C and 400 °C, greater hardening was found in the Fe-Cr alloys irradiated at the lower dose rate. Transmission electron microscopy and atom probe tomography of Fe 5%Cr identified larger dislocation loop densities and sizes in the alloy irradiated with the high dose rate and Cr precipitation in the alloy irradiated with the low dose rate. Loss of defects at extended sinks such as dislocations and grain boundaries. Growth or shrinkage of defect clusters by the capture of point defects. Mutual annihilation by the recombination of a vacancy and interstitial. At low dose rates and/or high irradiation temperatures, reaction path (i) (sinks) dominates and at a high dose rates and/or low irradiation temperature reaction path (iii) (recombination) dominates [2]. The evolution of radiation damage such as dislocation loops and voids and phenomena such as radiation induced segregation, swelling and creep, depend on the fraction of point defects which migrate to sinks, recombine or cluster within the lattice and will be influenced by the reaction path that dominates the microstructural evolution of the material under irradiation.The relative proportions of these reaction types are directly dependent on the density and mobility of the defects, and hence dependent on dose rate and temperature. In iron, vacancy type defects are generally found to have significantly higher activation energy for migration compared to interstitial

  16. Enhancing catalytic performance of palladium in gold and palladium alloy nanoparticles for organic synthesis reactions through visible light irradiation at ambient temperatures.

    PubMed

    Sarina, Sarina; Zhu, Huaiyong; Jaatinen, Esa; Xiao, Qi; Liu, Hongwei; Jia, Jianfeng; Chen, Chao; Zhao, Jian

    2013-04-17

    The intrinsic catalytic activity of palladium (Pd) is significantly enhanced in gold (Au)-Pd alloy nanoparticles (NPs) under visible light irradiation at ambient temperatures. The alloy NPs strongly absorb light and efficiently enhance the conversion of several reactions, including Suzuki-Miyaura cross coupling, oxidative addition of benzylamine, selective oxidation of aromatic alcohols to corresponding aldehydes and ketones, and phenol oxidation. The Au/Pd molar ratio of the alloy NPs has an important impact on performance of the catalysts since it determines both the electronic heterogeneity and the distribution of Pd sites at the NP surface, with these two factors playing key roles in the catalytic activity. Irradiating with light produces an even more profound enhancement in the catalytic performance of the NPs. For example, the best conversion rate achieved thermally at 30 °C for Suzuki-Miyaura cross coupling was 37% at a Au/Pd ratio of 1:1.86, while under light illumination the yield increased to 96% under the same conditions. The catalytic activity of the alloy NPs depends on the intensity and wavelength of incident light. Light absorption due to the Localized Surface Plasmon Resonance of gold nanocrystals plays an important role in enhancing catalyst performance. We believe that the conduction electrons of the NPs gain the light absorbed energy producing energetic electrons at the surface Pd sites, which enhances the sites' intrinsic catalytic ability. These findings provide useful guidelines for designing efficient catalysts composed of alloys of a plasmonic metal and a catalytically active transition metal for various organic syntheses driven by sunlight.

  17. Enhancing catalytic performance of palladium in gold and palladium alloy nanoparticles for organic synthesis reactions through visible light irradiation at ambient temperatures.

    PubMed

    Sarina, Sarina; Zhu, Huaiyong; Jaatinen, Esa; Xiao, Qi; Liu, Hongwei; Jia, Jianfeng; Chen, Chao; Zhao, Jian

    2013-04-17

    The intrinsic catalytic activity of palladium (Pd) is significantly enhanced in gold (Au)-Pd alloy nanoparticles (NPs) under visible light irradiation at ambient temperatures. The alloy NPs strongly absorb light and efficiently enhance the conversion of several reactions, including Suzuki-Miyaura cross coupling, oxidative addition of benzylamine, selective oxidation of aromatic alcohols to corresponding aldehydes and ketones, and phenol oxidation. The Au/Pd molar ratio of the alloy NPs has an important impact on performance of the catalysts since it determines both the electronic heterogeneity and the distribution of Pd sites at the NP surface, with these two factors playing key roles in the catalytic activity. Irradiating with light produces an even more profound enhancement in the catalytic performance of the NPs. For example, the best conversion rate achieved thermally at 30 °C for Suzuki-Miyaura cross coupling was 37% at a Au/Pd ratio of 1:1.86, while under light illumination the yield increased to 96% under the same conditions. The catalytic activity of the alloy NPs depends on the intensity and wavelength of incident light. Light absorption due to the Localized Surface Plasmon Resonance of gold nanocrystals plays an important role in enhancing catalyst performance. We believe that the conduction electrons of the NPs gain the light absorbed energy producing energetic electrons at the surface Pd sites, which enhances the sites' intrinsic catalytic ability. These findings provide useful guidelines for designing efficient catalysts composed of alloys of a plasmonic metal and a catalytically active transition metal for various organic syntheses driven by sunlight. PMID:23566035

  18. Release of helium from irradiation damage in Fe?9Cr ferritic alloy

    NASA Astrophysics Data System (ADS)

    Ono, K.; Arakawa, K.; Shibasaki, H.; Kurata, H.; Nakamichi, I.; Yoshida, N.

    2004-08-01

    Thermal desorption of helium from Fe-9Cr and pure Fe which were irradiated with 5 keV He + ions at 85 or 473 K have been studied by thermal desorption spectrometry (TDS) at temperatures from room temperature to 1270 K and related microstructure changes by TEM. It is found that five TDS peaks appear in Fe-9Cr specimen and these peaks are well correlated with microstructure changes. From these results, the five peaks I Cr, II Cr, III Cr, IV Cr and V Cr may be attributed to release of trapped helium atoms by the break-up of vacancy-helium-self-interstitial atom complexes or very tiny loops, glide to the specimen surface or coalescence of interstitial type dislocation loops, shrinkage of dislocation loops, bubble migration to the specimen surface and α-γ phase transformation, respectively. Precipitation of Cr around the loops and bubbles in Fe-9Cr is revealed by STEM-EELS and this could shift the related TDS peaks to higher temperatures.

  19. Formation and evolution of MnNi clusters in neutron irradiated dilute Fe alloys modelled by a first principle-based AKMC method

    NASA Astrophysics Data System (ADS)

    Ngayam-Happy, R.; Becquart, C. S.; Domain, C.; Malerba, L.

    2012-07-01

    An atomistic Monte Carlo model parameterised on electronic structure calculations data has been used to study the formation and evolution under irradiation of solute clusters in Fe-MnNi ternary and Fe-CuMnNi quaternary alloys. Two populations of solute rich clusters have been observed, which can be discriminated by whether or not the solute atoms are associated with self-interstitial clusters. Mn-Ni-rich clusters are observed at a very early stage of the irradiation in both modelled alloys, whereas the quaternary alloys contain also Cu-containing clusters. Mn-Ni-rich clusters nucleate very early via a self-interstitial-driven mechanism, earlier than Cu-rich clusters; the latter, however, which are likely to form via a vacancy-driven mechanism, grow in number much faster than the former, helped by the thermodynamic driving force to Cu precipitation in Fe, thereby becoming dominant in the low dose regime. The kinetics of the number density increase of the two populations is thus significantly different. Finally the main conclusion suggested by this work is that the so-called late blooming phases might as well be neither late, nor phases.

  20. Production and Characterization of Atomized U-Mo Powder by the Rotating Electrode Process

    SciTech Connect

    C.R. Clark; B.R. Muntifering; J.F. Jue

    2007-09-01

    In order to produce feedstock fuel powder for irradiation testing, the Idaho National Laboratory has produced a rotating electrode type atomizer to fabricate uranium-molybdenum alloy fuel. Operating with the appropriate parameters, this laboratory-scale atomizer produces fuel in the desired size range for the RERTR dispersion experiments. Analysis of the powder shows a homogenous, rapidly solidified microstructure with fine equiaxed grains. This powder has been used to produce irradiation experiments to further test adjusted matrix U-Mo dispersion fuel.

  1. Microstructural Characterization of U-Nb-Zr, U-Mo-Nb, and U-Mo-Ti Alloys via Electron Microscopy

    SciTech Connect

    A. Ewh; D. D. Keiser, Jr.; Y. H. Sohn

    2010-06-01

    Ternary uranium molybdenum alloys are currently being investigated for use as dispersion and monolithic nuclear fuels in research reactors. In this study, two such ternary alloys, with compositions U-8Mo-3Nb and U-7Mo-3Ti in wt.%, were examined using scanning electron microscopy (SEM), X-ray diffraction (XRD), and transmission electron microscopy (TEM) with high angle annular dark field (HAADF) imaging via scanning transmission electron microscopy (STEM) to identify phase constituents. These alloys were homogenized at 950°C for 96 hours and were expected to be single-phase bcc-!-U. However, upon examination, it was determined that despite homogenization, each of the alloys contained a small volume fraction precipitate phase. Through SEM and XRD, it was confirmed that the matrix retained the bcc-!-U phase, but the precipitate phases could not be identified using these methods. TEM specimens were prepared using site-specific focused ion beam (FIB) in situ lift out (INLO) technique to include at least one precipitate from each alloy. By electron diffraction, the precipitate phases for the U- 8Mo-3Nb and U-7Mo-3Ti alloys were identified as bcc-(Mo,Nb) solid solution and bcc- (Mo,Ti) solid solution, respectively.

  2. Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors

    NASA Astrophysics Data System (ADS)

    Kotlyarevskiy, S. G.; Pavliuk, A. O.; Zakharova, E. V.; Volkova, A. G.

    2016-06-01

    The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products: 93mNb and the long-lived 94Nb, 93Zr radionuclides. Radionuclides of 60Co, 137Cs, 90Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for 60Co and 137Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.

  3. Characterisation of Cr, Si and P distribution at dislocations and grain-boundaries in neutron irradiated Fe-Cr model alloys of low purity

    NASA Astrophysics Data System (ADS)

    Kuksenko, V.; Pareige, C.; Genevois, C.; Pareige, P.

    2013-03-01

    Segregations at some dislocations and grain boundaries in Fe-5%Cr, Fe-9%Cr and Fe-12%Cr model alloys of low purity after neutron irradiation at 300 °C up to 0.6 dpa have been analyzed with atom probe tomography. All dislocation lines and low- and high-angle grain boundaries (GBs) which have been observed were enriched with Cr, Si and P. The segregations reveal the different dislocation structures associated to different type of analysed GBs. Cr and Si atoms were found to be nonhomogenously distributed around the dislocation cores because of the non isotropic stress field induced by edge dislocation lines. Concerning GBs, precipitate free zones (PFZs) are exhibited around the planar defects which were analysed in Fe-9%Cr and Fe-12%Cr model alloys. These PFZ are size dependant with the nominal level of Cr.

  4. Specification of CuCrZr Alloy Properties after Various Thermo-Mechanical Treatments and Design Allowables including Neutron Irradiation Effects

    SciTech Connect

    Barabash, Vladimir; Kalinin, G. M.; Fabritsiev, Sergei A.; Zinkle, Steven J

    2012-01-01

    Precipitation hardened CuCrZr alloy is a promising heat sink and functional material for various applica- tions in ITER, for example the first wall, blanket electrical attachment, divertor, and heating systems. Three types of thermo-mechanical treatment were identified as most promising for the various applica- tions in ITER: solution annealing, cold working and ageing; solution annealing and ageing; solution annealing and ageing at non-optimal condition due to specific manufacturing processes for engineer- ing-scale components. The available data for these three types of treatments were assessed and mini- mum tensile properties were determined based on recommendation of Structural Design Criteria for the ITER In-vessel Components. The available data for these heat treatments were analyzed for assess- ment of neutron irradiation effect. Using the definitions of the ITER Structural Design Criteria the design allowable stress intensity values are proposed for CuCrZr alloy after various heat treatments.

  5. Microstructural development in waste form alloys cast from irradiated cladding residual from the electrometallurgical treatment of EBR-II spent fuel.

    SciTech Connect

    Keiser, D. D., Jr.

    1999-06-10

    A metallic waste form alloy that consists primarily of stainless steel and zirconium is being developed by Argonne National Laboratory to contain metallic waste constituents that are residual from an electrometallurgical treatment process for spent nuclear fuel. Ingots have been cast in an induction furnace in a hot cell using actual, leftover, irradiated, EBR-II cladding hulls treated in an electrorefiner. The as-cast ingots have been sampled using a core-drilling and an injection-casting technique. In turn, generated samples have been characterized using chemical analysis techniques and a scanning electron microscope equipped with energy dispersive and wavelength-dispersive spectrometers. As-cast ingots contain the predicted concentration levels of the various constituents, and most of the phases that develop are analogous to those for alloys generated using non-radioactive surrogates for the various fission products.

  6. Unusual response of the binary V-2Si alloy to neutron irradiation in FFTF at 430-600{degrees}C

    SciTech Connect

    Ohnuki, S.; Konoshita, H.; Takahaski, H.; Garner, F.A.

    1996-04-01

    When V-2Si was irradiated in FFTF at 430, 500 and 600C to doses as high as 80 dpa, a very unusual swelling response was observed in which the swelling appeared to saturate rather quickly at {approx}35% at 430 and 540C, but approached this swelling same level much more slowly at 600C. The possible causes of this phenomenon are discussed as well as the implications of these findings on the swelling behavior of other high swelling vanadium binary alloys.

  7. Ballistic effects on the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-Cu alloy.

    PubMed

    Xu, Donghua; Certain, Alicia; Lee Voigt, Hyon-Jee; Allen, Todd; Wirth, Brian D

    2016-09-14

    Studies of solute precipitation and precipitate phase stability in nuclear structural materials under concurrent irradiation and heat often lead to contradictory results due to the complex nature of the phenomena which is far from well understood. Here, we present a comprehensive atomistically based continuum model for the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-0.78 at. % Cu alloy. Our model incorporates thermal and irradiation enhanced diffusion of atomic Cu, clustering of Cu into sub-nanometer and nanometer sized precipitates, thermal dissociation of the precipitates and, in particular, a cascade re-dissolution parameter that has been made available by recent molecular dynamics simulations. Our model suggests that the Cu precipitates may form, re-dissolve, or coarsen under different irradiation and thermal conditions depending on the competition between the thermal and the ballistic effects. The quantitative predictions of our model are compared with available experiments including limited atom probe tomography data acquired in this study. The work highlights the importance of combining thermal and ballistic effects in the understanding of phase stability in extreme nuclear environments. PMID:27634272

  8. Ballistic effects on the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-Cu alloy

    NASA Astrophysics Data System (ADS)

    Xu, Donghua; Certain, Alicia; Lee Voigt, Hyon-Jee; Allen, Todd; Wirth, Brian D.

    2016-09-01

    Studies of solute precipitation and precipitate phase stability in nuclear structural materials under concurrent irradiation and heat often lead to contradictory results due to the complex nature of the phenomena which is far from well understood. Here, we present a comprehensive atomistically based continuum model for the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-0.78 at. % Cu alloy. Our model incorporates thermal and irradiation enhanced diffusion of atomic Cu, clustering of Cu into sub-nanometer and nanometer sized precipitates, thermal dissociation of the precipitates and, in particular, a cascade re-dissolution parameter that has been made available by recent molecular dynamics simulations. Our model suggests that the Cu precipitates may form, re-dissolve, or coarsen under different irradiation and thermal conditions depending on the competition between the thermal and the ballistic effects. The quantitative predictions of our model are compared with available experiments including limited atom probe tomography data acquired in this study. The work highlights the importance of combining thermal and ballistic effects in the understanding of phase stability in extreme nuclear environments.

  9. COMPOSITIONAL AND TEMPERATURE DEPENDENCE OF VOID SWELLING IN MODEL Fe-Cr BASE ALLOYS IRRADIATED IN THE EBR-II FAST REACTOR

    SciTech Connect

    Sencer, Bulent H.; Garner, Francis A.

    2001-04-01

    A series of annealed and aged Fe-xCr, Fe-12Cr-yC and Fe-12Cr-0.1C-zMo model alloys were irradiated in EBR-II at eight temperatures between 400 and 650C and dose levels ranging from 35 to 131 dpa. Swelling-induced density changes observed in the binary alloys generally peaked at mid-chromium levels, with the chromium and temperature dependence expressed primarily in the duration of the transient regime. The steady state swelling rate at the lower irradiation temperatures was much higher than previous estimates, reaching ~0.2%/dpa and possibly still climbing at higher neutron exposures. The dependence of swelling on molybdenum and carbon was more complex, depending on whether the temperature was relatively low or high. At temperatures of 482oC and above the effect of carbon additions was very pronounced, with swelling of Fe-12Cr jumping dramatically from near zero at 0.002%C to 6-10% at 0.1%C. This indicates that the major determinant of the composition and temperature dependence probably lies in the duration of the nucleation-dominated transient regime of swelling and not primarily in the steady-state swelling rate as previously envisioned. This raises the possibility that significant swelling may occur earlier in fusion and spallation neutron spectra where high gas generation rates may assist void nucleation.

  10. Roles of Vacancy/Interstitial Diffusion and Segregation in the Microchemistry at Grain Boundaries of Irradiated Fe-Cr-Ni alloys

    DOE PAGES

    Yang, Ying; Field, Kevin G.; Allen, Todd R.; Busby, Jeremy T.

    2016-02-23

    A detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe–Cr–Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling, is presented. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. The calculated fluxes up to 10 dpa suggested the dominant diffusion mechanism for chromium and iron is via vacancy,more » while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.« less

  11. Effect of electron irradiation on the 3C-4H transformation in a Co-Fe alloy

    SciTech Connect

    Allen, C.W.; Mori, H.

    1996-03-01

    Preliminary results are presented of a study of electron irradiation on the martensitic transformation in a relatively dilute terminal solid solution of Co+5.75 wt% Fe. The experiments demonstrate qualitatively the profound effect that the damage structure produced by electron irradiation can have on these simple martensitic transformations. Further study is being pursued to attempt to quantify the transformation retardation effect.

  12. Ion mass dependence of irradiation-induced local creation of ferromagnetism in Fe{sub 60}Al{sub 40} alloys

    SciTech Connect

    Fassbender, J.; Liedke, M. O.; Strache, T.; Moeller, W.; Menendez, E.; Sort, J.; Rao, K. V.; Deevi, S. C.; Nogues, J.

    2008-05-01

    Ion irradiation of Fe{sub 60}Al{sub 40} alloys results in the phase transformation from the paramagnetic, chemically ordered B2 phase to the ferromagnetic, chemically disordered A2 phase. The magnetic phase transformation is related to the number of displacements per atom (dpa) during the irradiation. For heavy ions (Ar{sup +}, Kr{sup +}, and Xe{sup +}), a universal curve is observed with a steep increase in the fraction of the ferromagnetic phase that reaches saturation, i.e., a complete phase transformation, at about 0.5 dpa. This proves the purely ballistic nature of the disordering process. If light ions are used (He{sup +} and Ne{sup +}), a pronounced deviation from the universal curve is observed. This is attributed to bulk vacancy diffusion from the dilute collision cascades, which leads to a partial recovery of the thermodynamically favored B2 phase. Comparing different noble gas ion irradiation experiments allows us to assess the corresponding counteracting contributions. In addition, the potential to create local ferromagnetic areas embedded in a paramagnetic matrix is demonstrated.

  13. Tensile properties of vanadium-base alloys irradiated in the Fusion-1 low-temperature experiment in the BOR-60 reactor

    SciTech Connect

    Tsai, H.; Gazda, J.; Nowicki, L.J.; Billone, M.C.; Smith, D.L.

    1998-09-01

    The irradiation has been completed and the test specimens have been retrieved from the lithium-bonded capsule at the Research Institute of Atomic Reactors (RIAR) in Russia. During this reporting period, the Argonne National Laboratory (ANL) tensile specimens were received from RIAR and initial testing and examination of these specimens at ANL has been completed. The results, corroborating previous findings showed a significant loss of work hardening capability in the materials. There appears to be no significant difference in behavior among the various heats of vanadium-base alloys in the V-(4-5)Cr-(4-5)Ti composition range. The variations in the preirradiation annealing conditions also produced no notable differences.

  14. Irradiation effects on 17-7 PH stainless steel, A-201 carbon steel, and titanium-6-percent-aluminum-4-percent-vanadium alloy

    NASA Technical Reports Server (NTRS)

    Hasse, R. A.; Hartley, C. B.

    1972-01-01

    Irradiation effects on three materials from the NASA Plum Brook Reactor Surveillance Program were determined. An increase of 105 K in the nil-ductility temperature for A-201 steel was observed at a fluence of approximately 3.1 x 10 to the 18th power neutrons/sq cm (neutron energy E sub n greater than 1.0 MeV). Only minor changes in the mechanical properties of 17-7 PH stainless steel were observed up to a fluence of 2 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV). The titanium-6-percent-aluminum-4-percent-vanadium alloy maintained its notch toughness up to a fluence of 1 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV).

  15. Scattering effects and high-spatial-frequency nanostructures on ultrafast laser irradiated surfaces of zirconium metallic alloys with nano-scaled topographies.

    PubMed

    Li, Chen; Cheng, Guanghua; Sedao, Xxx; Zhang, Wei; Zhang, Hao; Faure, Nicolas; Jamon, Damien; Colombier, Jean-Philippe; Stoian, Razvan

    2016-05-30

    The origin of high-spatial-frequency laser-induced periodic surface structures (HSFL) driven by incident ultrafast laser fields, with their ability to achieve structure resolutions below λ/2, is often obscured by the overlap with regular ripples patterns at quasi-wavelength periodicities. We experimentally demonstrate here employing defined surface topographies that these structures are intrinsically related to surface roughness in the nano-scale domain. Using Zr-based bulk metallic glass (Zr-BMG) and its crystalline alloy (Zr-CA) counterpart formed by thermal annealing from its glassy precursor, we prepared surfaces showing either smooth appearances on thermoplastic BMG or high-density nano-protuberances from randomly distributed embedded nano-crystallites with average sizes below 200 nm on the recrystallized alloy. Upon ultrashort pulse irradiation employing linearly polarized 50 fs, 800 nm laser pulses, the surfaces show a range of nanoscale organized features. The change of topology was then followed under multiple pulse irradiation at fluences around and below the single pulse threshold. While the former material (Zr-BMG) shows a specific high quality arrangement of standard ripples around the laser wavelength, the latter (Zr-CA) demonstrates strong predisposition to form high spatial frequency rippled structures (HSFL). We discuss electromagnetic scenarios assisting their formation based on near-field interaction between particles and field-enhancement leading to structure linear growth. Finite-difference-time-domain simulations outline individual and collective effects of nanoparticles on electromagnetic energy modulation and the feedback processes in the formation of HSFL structures with correlation to regular ripples (LSFL). PMID:27410083

  16. Formation of microcraters and hierarchically-organized surface structures in TiNi shape memory alloy irradiated with a low-energy, high-current electron beam

    SciTech Connect

    Meisner, L. L. Meisner, S. N.; Markov, A. B. Ozur, G. E. Yakovlev, E. V.; Rotshtein, V. P.; Gudimova, E. Yu.

    2015-10-27

    The regularities of surface cratering in TiNi alloy irradiated with a low-energy, high-current electron beam (LEHCEB) in dependence on energy density and number of pulses are studied. LEHCEB processing of TiNi samples was carried out using RITM-SP facility. Energy density E{sub s} was varied from 1 to 5 J/cm{sup 2}, pulse duration was 2.5–3.0 μs, the number of pulses n = 1–128. The dominant role of non-metallic inclusions [mainly, TiC(O)] in the nucleation of microcraters was found. It was revealed that at small number of pulses (n = 2), an increase in energy density leads both to increasing average diameter and density of microcraters. An increase in the number of pulses leads to a monotonic decrease in density of microcraters, and, therefore, that of the proportion of the area occupied by microcraters, as well as a decrease in the surface roughness. The multiple LEHCEB melting of TiNi alloy in crater-free modes enables to form quasi-periodical, hierarchically-organized microsized surface structures.

  17. Effects of pulsed dual-ion irradiation on phase transformations and microstructure in Ti-modified austenitic alloy

    SciTech Connect

    Lee, E.H.; Packan, N.H.; Mansur, L.K.

    1983-01-05

    The influence of pulsed 4 MeV Ni ion bombardment, with and without simultaneous helium injection, has been explored in a low swelling, Ti-modified austenitic stainless steel. Irradiations were carried out to 70 dpa at 950/sup 0/K; the pulsing frequencies were either 60 s on/off or 1 s on/off. Compared to continuous irradiation, pulsing caused a decrease in the interstitial loop diameter at 1 dpa, although at higher doses the overall dislocation density was not affected. Pulsing and helium both promoted the stability of MC precipitates and retarded the subsequent G phase formation; in some cases G-phase was suppressed and eta phase formed instead. Small bubble-like cavities were observed to grow into large voids after steady dual beam irradiation to 70 dpa. However, this conversion was suppressed by pulse irradiation to 70 dpa and furthermore the sizes of the small cavities were somewhat reduced. The results are explained in terms of current mechanistic understanding of mean point defect kinetics and the evolution of microstructure and microcomposition during irradiation with superimposed annealing periods.

  18. Evaluation of ferritic alloy Fe-2-1/4Cr-1Mo after neutron irradiation - microstructure development

    SciTech Connect

    Gelles, D.S.

    1984-05-01

    Microstructural examinations are reported for nine specimen conditions of 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510/sup 0/C. Two heats of material were involved, each with a different preirradiation heat treatment, one irradiated to a peak fluence of 5.1 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) or 24 dpa and the other to 2.4 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV) or 116 dpa. Void swelling is found following irradiation at 400/sup 0/C in both conditions and to 480/sup 0/C in the higher fluence conditions. Concurrently dislocation structure and precipitation formed. Peak void swelling, void density, dislocation density and precipitate number density developed at the lowest temperature, approx. 400/sup 0/C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior.

  19. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  20. Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

    SciTech Connect

    Ozaltun, Hakan; Medvedev, Pavel G

    2015-06-01

    Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.

  1. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    DOE PAGES

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms ofmore » the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.« less

  2. Low dose of continuous – wave microwave irradiation did not cause temperature increase in muscles tissue adjacent to titanium alloy implants – an animal study

    PubMed Central

    2013-01-01

    Background Research studies on the influence of radiofrequency electromagnetic radiation on implants in vitro have failed to investigate temperature changes in the tissues adjacent to the implants under microwave therapy. We therefore, used a rabbit model in an effort to determine the impact of microwave therapy on temperature changes in tissues adjacent to the titanium alloy implants and the safety profile thereof. Methods Titanium alloy internal fixation plates were implanted in New Zealand rabbits in the middle of femur. Microwave therapy was performed by a 2450 MHz microwave generator 3 days after the surgery. Temperature changes of muscles adjacent to the implants were recorded under exposure to dose-gradient microwave radiation from 20w to 60w. Results Significant difference between control and microwave treatment group at peak temperatures (Tpeak) and temperature gap (Tgap= Tpeak-Tvally) were observed in deep muscles (Tpeak, 41.63 ± 0.21°C vs. 44.40 ± 0.17°C, P < 0.01; Tgap, 5.33 ± 0.21°C vs. 8.10 ± 0.36°C, P < 0.01) and superficial muscles (Tpeak, 41.53 ± 0.15°C vs. 42.03 ± 0.23°C, P = 0.04; Tgap, 5.23 ± 0.21°C vs. 5.80 ± 0.17°C, P = 0.013) under 60 w, and deep muscles (Tpeak, 40.93 ± 0.25°C vs. 41.87 ± 0.23°C, P = 0.01; Tgap, 4.73 ± 0.20°C vs. 5.63 ± 0.35°C, P = 0.037) under 50w, but not under 20, 30 and 40w. Conclusion Our results suggest that low-dose (20w-40w) continuous-wave microwave irradiation delivered by a 2450 MHz microwave generator might be a promising treatment for patients with titanium alloy internal fixation, as it did not raise temperature in muscle tissues adjacent to the titanium alloy implant. PMID:24365389

  3. The measurement of material degradation in neutron irradiated Mn-Mo-Ni low alloy steels using magnetic techniques

    NASA Astrophysics Data System (ADS)

    Chang, Kee-Ok; Chi, Se-Hwan; Kim, Taek-Soo; Kim, Byoung-Chul; Lee, Sam-Lai

    1999-12-01

    To examine the application of magnetic method to the evaluation of radiation damage and thermal recovery in Mn-Mo-Ni reactor pressure vessel(RPV) steels, changes in the magnetic parameters and Vickers microhardness(Hv) due to neutron irradiation and heat treatment were measured and compared for RPV surveillance specimens which were irradiated to the neutron fluence of 2.4×1019n/cm2 (E⩾1.0 MeV) in a typical pressurized water reactor environment at 288 °C. Results show that the coercivity(Hc) increased due to irradiation, whereas susceptibility and Barkhausen noise energy(BNE) decreased. Saturation magnetization(Ms) remained unchanged while Vickers microhardness increased. For isothermally heat treated specimens, both the magnetic parameters and Vickers microhardness returned to the unirradiated condition: Thus, the BNE and susceptibility increased while the coercivity and Vickers microhardness decreased. From the comparison of irradiated base metal and weld metal, it is found that the susceptibility and BNE of the weld metal were smaller than those of base metal and the microhardness and coercivity of weld metal were larger than those of base metal. All these consistent changes in magnetic parameters show a possibility that magnetic techniques may be used for the evaluation of material degradation.

  4. Aberration-corrected X-ray spectrum imaging and Fresnel contrast to differentiate nanoclusters and cavities in helium-irradiated alloy 14YWT

    SciTech Connect

    Miller, Michael K; Parish, Chad M

    2014-01-01

    Helium accumulation negatively impacts structural materials used in neutron-irradiated environments, such as fission and fusion reactors. Next-generation fission and fusion reactors will require structural materials, such as steels, resistant to large neutron doses yet see service temperatures in the range most affected by helium embrittlement. Previous work has indicated the difficulty of experimentally differentiating nanometer-sized helium bubbles from the Ti-Y-O rich nanoclustsers (NCs) in radiation-tolerant nanostructured ferritic alloys (NFAs). Because the NCs are expected to sequester helium away from grain boundaries and reduce embrittlement, experimental methods to study simultaneously the NC and bubble populations are needed. In this study, aberration-corrected scanning transmission electron microscopy (STEM) results combining high-collection-efficiency X-ray spectrum images (SIs), multivariate statistical analysis (MVSA), and Fresnel-contrast bright-field STEM imaging have been used for such a purpose. Results indicate that Fresnel-contrast imaging, with careful attention to TEM-STEM reciprocity, differentiates bubbles from NCs, and MVSA of X-ray SIs unambiguously identifies NCs. Therefore, combined Fresnel-contrast STEM and X-ray SI is an effective STEM-based method to characterize helium-bearing NFAs.

  5. MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

    SciTech Connect

    Farrell, K.

    2003-09-26

    Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components. These three alloys cover the range of crystal structures usually encountered in structural alloys, i.e. body-centered cubic (bcc), face-centered cubic (fcc), and close-packed hexagonal (cph), respectively. The experiments were conducted in three Phases, corresponding to the three years duration of the project. Phases 1 and 2 addressed irradiations and tensile tests made at near-ambient temperatures, and covered a wide range of neutron fluences

  6. Cleavage fracture and irradiation embrittlement of fusion reactor alloys: mechanisms, multiscale models, toughness measurements and implications to structural integrity assessment

    NASA Astrophysics Data System (ADS)

    Odette, G. R.; Yamamoto, T.; Rathbun, H. J.; He, M. Y.; Hribernik, M. L.; Rensman, J. W.

    2003-12-01

    We describe the highly efficient master curves-shifts (MC-Δ T) method to measure and apply cleavage fracture toughness, KJc ( T), data and show that it is applicable to 9Cr martensitic steels. A reference temperature, T0, indexes the invariant MC shape on an absolute temperature scale. Then, T0 shifts (Δ T) are used to account for various effects of size and geometry, loading rate and irradiation embrittlement (Δ Ti). The paper outlines a multiscale model, relating atomic to structural scale fracture processes, that underpins the MC-Δ T method. At the atomic scale, we propose that the intrinsic microarrest toughness, Kμ( T), of the body-centered cubic ferrite lattice dictates an invariant shape of the macroscopic KJc ( T) curve. KJc ( T) can be modeled in terms of the true stress-strain ( σ- ɛ) constitutive law, σ ( T, ɛ), combined with a temperature-dependent critical local stress, σ*( T) and stressed volume, V*. The local fracture properties, σ*( T)- V*, are governed by coarse-scale brittle trigger particles and Kμ( T). Irradiation (and high strain rate) induced increases in the yield stress, Δ σy, lead to Δ Ti, with typical Δ Ti/Δ σy≈0.6±0.15 °C/MPa. However, Δ Ti associated with decreases in σ* and V* can result from a number of potential non-hardening embrittlement (NHE) mechanisms, including a large amount of He on grain boundaries. Estimates based on available data suggest that this occurs at >500-700 appm bulk He. Hardening and NHE are synergistic, and can lead to very large Δ Ti. NHE is signaled by large (>1 °C/MPa), or even negative, values of Δ Ti/Δ σy (for Δ σy<0), and is often coupled with increasing amounts of intergranular fracture. The measured and effective fracture toughness pertinent to structures almost always depends on the size and geometry of the cracked body, and is typically significantly greater than KJc . Size and geometry effects arise from both weakest link statistics, related to the volume under high

  7. Acceleration of ordering transformation of a new Fe{sub 2}(Mn,Cr)Si Heusler-alloy film by very high frequency plasma irradiation process during radio frequency sputter deposition

    SciTech Connect

    Yoshimura, S.; Kobayashi, H.; Egawa, G.; Saito, H.; Ishida, S.

    2011-04-01

    A new Heusler alloy, Fe{sub 2}(Mn,Cr)Si, that is likely to have high spin polarization (P) and high damping constant ({alpha}) was proposed to obtain high magneto-resistance ratio and low spin torque noise in a magnetic read head with a current-perpendicular-to-plane (CPP) giant magneto-resistance (GMR) multilayer. A very high frequency (VHF) plasma irradiation process during radio frequency (RF) sputter deposition was investigated to form the highly ordered structure of the Heusler alloy film with low thermal treatment temperature. The main results are as follows: (1) P and magnetic moment of Fe{sub 2}(Mn{sub 0.5}Cr{sub 0.5})Si with an L2{sub 1} structure were estimated at 0.99 and 2.49 {mu}{sub B}/f.u., respectively, and {alpha} was also estimated to be larger compared with the case of Co{sub 2}MnSi, according to density of states (DOS) calculations. (2) The ordering (at least B2 structure) temperature of Fe{sub 2}(Mn{sub 0.6}Cr{sub 0.4})Si film decreased from 500 to 300 deg. C by using the VHF plasma irradiation process with optimized condition. (3) The surface roughness of Fe{sub 2}(Mn{sub 0.6}Cr{sub 0.4})Si film also reduced from 1.7 to 0.5 nm by using the VHF plasma irradiation process. It is found that the Fe{sub 2}(Mn,Cr)Si Heusler alloy and the VHF plasma irradiation process with optimized condition seems to be applicable for fabrication of high-performance magnetic read head with CPP-GMR device.

  8. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    SciTech Connect

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  9. Radiation Effects in Refractory Alloys

    NASA Astrophysics Data System (ADS)

    Zinkle, Steven J.; Wiffen, F. W.

    2004-02-01

    In order to achieve the required low reactor mass per unit electrical power for space reactors, refractory alloys are essential due to their high operating temperature capability that in turn enables high thermal conversion efficiencies. One of the key issues associated with refractory alloys is their performance in a neutron irradiation environment. The available radiation effects data are reviewed for alloys based on Mo, W, Re, Nb and Ta. The largest database is associated with Mo alloys, whereas Re, W and Ta alloys have the least available information. Particular attention is focused on Nb-1Zr, which is a proposed cladding and structural material for the reactor in the Jupiter Icy Moons Orbiter (JIMO) project. All of the refractory alloys exhibit qualitatively similar temperature-dependent behavior. At low temperatures up to ~0.3TM, where TM is the melting temperature, the dominant effect of radiation is to produce pronounced radiation hardening and concomitant loss of ductility. The radiation hardening also causes a dramatic decrease in the fracture toughness of the refractory alloys. These low temperature radiation effects occur at relatively low damage levels of ~0.1 displacement per atom, dpa (~2×1024 n/m2, E>0.1 MeV). As a consequence, operation at low temperatures in the presence of neutron irradiation must be avoided for all refractory alloys. At intermediate temperatures (0.3 to 0.6 TM), void swelling and irradiation creep are the dominant effects of irradiation. The amount of volumetric swelling associated with void formation in refractory alloys is generally within engineering design limits (<5%) even for high neutron exposures (>>10 dpa). Very little experimental data exist on irradiation creep of refractory alloys, but data for other body centered cubic alloys suggest that the irradiation creep will produce negligible deformation for near-term space reactor applications.

  10. The Accumulation and Annealing of Radiation-Induced Defects and the Effect of Hydrogen on the Physicomechanical Properties of the V-4Ti-4Cr and V-10Ti-5Cr Vanadium-Based Alloys under Low-Temperature (at 77 K) Neutron Irradiation

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Zuev, Yu. N.; Kar'kin, A. E.; Parkhomenko, V. D.; Kozlov, A. V.; Chernov, V. M.

    2016-03-01

    The processes of the accumulation and annealing of radiation-induced defects that occur under low-temperature (at 77 K) irradiation (with an energy E > 0.1 MeV) of V-4Ti-4Cr and V-10Ti-5Cr bcc alloys both nonmodified and modified with hydrogen isotopes in a concentration of 200 ppm, as well as the effect of these processes on the physicomechanical properties of these alloys, have been studied. It has been found that the saturation of these alloys with hydrogen leads to slight changes in their strength and ductility characteristics. The irradiation of the alloys at the temperature of 77 K results in a substantial increase in their yield stress and ultimate strength, as well as a decrease in their ductility. In the course of the postradiation annealing of the alloys at a temperature of 130 K, the stage related to the migration of interstitial atoms is observed. At temperatures of 290-320 K, the recovery stage occurs due to the formation of vacancy clusters. The stage that occurs at a temperature of 470 K can be attributed to the formation of impurity-vacancy clusters. Possible mechanisms of the radiation-induced strengthening of the alloys during irradiation and subsequent annealing have been discussed.

  11. A positron beam Doppler broadening analysis of formation and recovery of defects produced by ion irradiation in Fe‒C‒Cu alloys

    NASA Astrophysics Data System (ADS)

    Iwai, Takeo

    2013-04-01

    Mechanisms of radiation embrittlement of reactor pressure vessel steels remain to be fully understood, particularly the nature of so-called 'matrix defects'. One possible mechanism is vacancy cluster formation, probably assisted by cascade damage. In order to investigate the effect of copper on the formation and annealing processes of vacancy clusters, ion-irradiated Fe‒C and Fe‒C‒Cu were investigated using a variable energy positron beam. Doppler broadening analysis revealed that vacancy-type defects are produced by ion irradiation and that copper addition reduces the open volume of the defects. Post irradiation annealing suggested the vacancy clusters do not have a substantial role in irradiation hardening.

  12. A master curve-mechanism based approach to modeling the effects of constraint, loading rate and irradiation on the toughness-temperature behavior of a V-4Cr-4Ti alloy

    SciTech Connect

    Odette, G.R.; Donahue, E.; Lucas, G.E.; Sheckherd, J.W.

    1996-10-01

    The influence of loading rate and constraint on the effective fracture toughness as a function of temperature [K{sub e}(T)] of the fusion program heat of V-4Cr-4Ti was measured using subsized, three point bend specimens. The constitutive behavior was characterized as a function of temperature and strain rate using small tensile specimens. Data in the literature on this alloy was also analysed to determine the effect of irradiation on K{sub e}(T) and the energy temperature (E-T) curves measured in subsized Charpy V-notch tests. It was found that V-4Cr-4Ti undergoes {open_quotes}normal{close_quotes} stress-controlled cleavage fracture below a temperature marking a sharp ductile-to-brittle transition. The transition temperature is increased by higher loading rates, irradiation hardening and triaxial constraint. Shifts in a reference transition temperature due to higher loading rates and irradiation can be reasonably predicted by a simple equivalent yield stress model. These results also suggest that size and geometry effects, which mediate constraint, can be modeled by combining local critical stressed area {sigma}*/A* fracture criteria with finite element method simulations of crack tip stress fields. The fundamental understanding reflected in these models will be needed to develop K{sub e}(T) curves for a range of loading rates, irradiation conditions, structural size scales and geometries relying (in large part) on small specimen tests. Indeed, it may be possible to develop a master K{sub e}(T) curve-shift method to account for these variables. Such reliable and flexible failure assessment methods are critical to the design and safe operation of defect tolerant vanadium structures.

  13. Preliminary report on the irradiation conditions of the HFIR JP-23 experiment

    SciTech Connect

    Ermi, A.M.; Gelles, D.S.

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

  14. Update on US High Density Fuel Fabrication Development

    SciTech Connect

    C.R. Clark; G.A. Moore; J.F. Jue; B.H. Park; N.P. Hallinan; D.M. Wachs; D.E. Burkes

    2007-03-01

    Second generation uranium molybdenum fuel has shown excellent in-reactor irradiation performance. This metallic fuel type is capable of being fabricated at much higher loadings than any presently used research reactor fuel. Due to the broad range of fuel types this alloy system encompasses—fuel powder to monolithic foil and binary fuel systems to multiple element additions—significant amounts of research and development have been conducted on the fabrication of these fuels. This paper presents an update of the US RERTR effort to develop fabrication techniques and the fabrication methods used for the RERTR-9A miniplate test.

  15. Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

    SciTech Connect

    Zinkle, S.J.; Alexander, D.J.; Robertson, J.P.

    1997-04-01

    Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimens were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.

  16. Scanning Electron Microscopy Analysis of Fuel/Matrix Interaction Layers in Highly-Irradiated U–Mo Dispersion Fuel Plates with Al and Al–Si Alloy Matrices

    SciTech Connect

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Brandon D. Miller; Jian Gan; Adam B. Robinson; Pavel Medvedev; James Madden; Dan Wachs; Mitch Meyer

    2014-04-01

    In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U–7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U–7Mo dispersion fuel elements with pure Al, Al–2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U–7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission-gas bubbles. Additionally, solid-fission-product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U–7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al–Si matrices.

  17. Swelling of Uranium Alloys at High Exposures

    SciTech Connect

    McDonell, W.R.

    2001-03-26

    This reports summarizes the results of postirradiation examinations of a series of unrestrained dilute uranium alloy specimens irradiated to exposures up to 13,000 MWD/T in NaK-containing stainless steel capsules.

  18. Alloy materials

    DOEpatents

    Hans Thieme, Cornelis Leo; Thompson, Elliott D.; Fritzemeier, Leslie G.; Cameron, Robert D.; Siegal, Edward J.

    2002-01-01

    An alloy that contains at least two metals and can be used as a substrate for a superconductor is disclosed. The alloy can contain an oxide former. The alloy can have a biaxial or cube texture. The substrate can be used in a multilayer superconductor, which can further include one or more buffer layers disposed between the substrate and the superconductor material. The alloys can be made a by process that involves first rolling the alloy then annealing the alloy. A relatively large volume percentage of the alloy can be formed of grains having a biaxial or cube texture.

  19. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Gardner, Levi D.; Casella, Andrew M.; Huber, Tanja K.; Breitkreutz, Harald

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  20. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2015-03-01

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  1. Modification of the titanium alloy surface in electroexplosive alloying with boron carbide and subsequent electron-beam treatment

    NASA Astrophysics Data System (ADS)

    Gromov, Victor E.; Budovskikh, Evgeniy A.; Ivanov, Yurii F.; Bashchenko, Lyudmila P.; Wang, Xinli; Kobzareva, Tatyana Yu.; Semin, Alexander P.

    2015-10-01

    The modification of the VT6 titanium alloy surface in electroexplosion alloying with plasma being formed in titanium foil with a weighed powder of boron carbide with subsequent irradiation by a pulsed electron beam has been carried out. An electroexplosive alloying zone of a thickness up to 50 μm with a gradient structure is found to form. The subsequent electron-beam treatment of the alloying zone results in smoothing of the alloying surface and is accompanied by the formation of the multilayer structure with alternating layers of various alloying degree at a depth of 30 μm.

  2. Modification of the titanium alloy surface in electroexplosive alloying with boron carbide and subsequent electron-beam treatment

    SciTech Connect

    Gromov, Victor E. Budovskikh, Evgeniy A. Bashchenko, Lyudmila P. Kobzareva, Tatyana Yu. Semin, Alexander P.; Ivanov, Yurii F.; Wang, Xinli

    2015-10-27

    The modification of the VT6 titanium alloy surface in electroexplosion alloying with plasma being formed in titanium foil with a weighed powder of boron carbide with subsequent irradiation by a pulsed electron beam has been carried out. An electroexplosive alloying zone of a thickness up to 50 μm with a gradient structure is found to form. The subsequent electron-beam treatment of the alloying zone results in smoothing of the alloying surface and is accompanied by the formation of the multilayer structure with alternating layers of various alloying degree at a depth of 30 μm.

  3. Experimental method for investigating helium effects in irradiated vanadium

    SciTech Connect

    Smith, D.L.; Matsui, H.; Greenwood, L.; Loomis, B.

    1987-10-01

    Analyses have been performed which indicate that an effective method for experimentally investigating helium effects in neutron irradiated vanadium base alloys can be developed. The experimental procedure involves only modest modifications to existing procedures currently used for irradiation testing of vanadium-base alloys in the FFTF reactor. Helium is generated in the vanadium alloy by decay of tritium which is either preinjected or generated within the test capsule. Calculations indicate that nearly constant He/dpa ratios of desired magnitude can be attained by proper selection of experimental parameters. The proposed method could have a major impact on the development of vanadium base alloys for fusion reactor applications. 8 refs., 4 figs.

  4. The nanostructure evolution in Fe-C systems under irradiation at 560 K

    NASA Astrophysics Data System (ADS)

    Jansson, V.; Chiapetto, M.; Malerba, L.

    2013-11-01

    This work extends our Object Kinetic Monte Carlo model for neutron irradiation-induced nanostructure evolution in Fe-C alloys to consider higher irradiation temperatures. The previous study concentrated on irradiation temperatures <370 K. Here we study the evolution of vacancy and self-interstitial atom (SIA) cluster populations at the operational temperature of light water reactors, by simulating specific reference irradiation experiments.

  5. ELECTRON IRRADIATION OF SOLIDS

    DOEpatents

    Damask, A.C.

    1959-11-01

    A method is presented for altering physical properties of certain solids, such as enhancing the usefulness of solids, in which atomic interchange occurs through a vacancy mechanism, electron irradiation, and temperature control. In a centain class of metals, alloys, and semiconductors, diffusion or displacement of atoms occurs through a vacancy mechanism, i.e., an atom can only move when there exists a vacant atomic or lattice site in an adjacent position. In the process of the invention highenergy electron irradiation produces additional vacancies in a solid over those normally occurring at a given temperature and allows diffusion of the component atoms of the solid to proceed at temperatures at which it would not occur under thermal means alone in any reasonable length of time. The invention offers a precise way to increase the number of vacancies and thereby, to a controlled degree, change the physical properties of some materials, such as resistivity or hardness.

  6. BRAZING ALLOYS

    DOEpatents

    Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

    1963-02-26

    A brazing alloy which, in the molten state, is characterized by excellent wettability and flowability, said alloy being capable of forming a corrosion resistant brazed joint wherein at least one component of said joint is graphite and the other component is a corrosion resistant refractory metal, said alloy consisting essentially of 20 to 50 per cent by weight of gold, 20 to 50 per cent by weight of nickel, and 15 to 45 per cent by weight of molybdenum. (AEC)

  7. VANADIUM ALLOYS

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1959-05-12

    This patent deals with vanadium based ternary alloys useful as fuel element jackets. According to the invention the ternary vanadium alloys, prepared in an arc furnace, contain from 2.5 to 15% by weight titanium and from 0.5 to 10% by weight niobium. Characteristics of these alloys are good thermal conductivity, low neutron capture cross section, good corrosion resistance, good welding and fabricating properties, low expansion coefficient, and high strength.

  8. The effect of solute on ultrasonic grain refinement of magnesium alloys

    NASA Astrophysics Data System (ADS)

    Qian, M.; Ramirez, A.; Das, A.; StJohn, D. H.

    2010-07-01

    An experimental study has been conducted to assess the structural refinement of magnesium and its alloys by ultrasonic irradiation during solidification. It is shown that (i) ultrasonic irradiation leads to significant refinement only in the presence of adequate solute, which is alloy dependent; (ii) the attendant grain density increases linearly with increase in solute content at a given irradiation level; (iii) increasing the solute content at a low irradiation level above the cavitation threshold is more effective than substantially increasing the irradiation intensity; and (iv) the difference in the grain size between two ultrasonicated magnesium alloys is mainly determined by the solute content rather than the irradiation intensity. In view of these, the effect of ultrasonic irradiation on solute redistribution in a solidifying magnesium alloy seems rather limited even at a substantial intensity level such as 1700 W cm -2. The implications of these findings are discussed and a mechanism is proposed to account for the experimental observations.

  9. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect

    Anderoglu, Osman; Tesmer, Joseph R.; Maloy, Stuart A.

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  10. URANIUM ALLOYS

    DOEpatents

    Seybolt, A.U.

    1958-04-15

    Uranium alloys containing from 0.1 to 10% by weight, but preferably at least 5%, of either zirconium, niobium, or molybdenum exhibit highly desirable nuclear and structural properties which may be improved by heating the alloy to about 900 d C for an extended period of time and then rapidly quenching it.

  11. ZIRCONIUM ALLOY

    DOEpatents

    Wilhelm, H.A.; Ames, D.P.

    1959-02-01

    A binary zirconiuin--antimony alloy is presented which is corrosion resistant and hard containing from 0.07% to 1.6% by weight of Sb. The alloys have good corrosion resistance and are useful in building equipment for the chemical industry.

  12. Nonswelling alloy

    DOEpatents

    Harkness, S.D.

    1975-12-23

    An aluminum alloy containing one weight percent copper has been found to be resistant to void formation and thus is useful in all nuclear applications which currently use aluminum or other aluminum alloys in reactor positions which are subjected to high neutron doses.

  13. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    SciTech Connect

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-02-25

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.

  14. Surface modification of Ti alloy by electro-explosive alloying and electron-beam treatment

    NASA Astrophysics Data System (ADS)

    Gromov, Victor; Kobzareva, Tatiana; Ivanov, Yuryi; Budovskikh, Evgeniy; Baschenko, Lyudmila

    2016-01-01

    By methods of modern physical metallurgy the analysis of structure phase states of titanium alloy VT6 is carried out after electric explosion alloying with boron carbide and subsequent irradiation by pulsed electron beam. The formation of an electro-explosive alloying zone of a thickness up to 50 µm, having a gradient structure, characterized by decrease in the concentration of carbon and boron with increasing distance to the treatable surface has been revealed. Subsequent electron-beam treatment of alloying zone leads to smoothing of the alloying area surface and is accompanied by the multilayer structure formation at the depth of 30 µm with alternating layers with different alloying degrees having the structure of submicro - and nanoscale level.

  15. Development and testing ov danadium alloys for fusion applications

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1996-10-01

    V base alloys have advantages for fusion reactor first-wall and blanket structure. To screen candidate alloys and optimize a V-base alloy, physical and mechanical properties of V-Ti, V-Cr-Ti, and V-Ti- Si alloys were studied before and after irradiation in Li environment in fast fission reactors. V-4Cr-4Ti containing 500-1000 wppM Si and <1000 wppM O+N+C was investigated as the most promising alloy, and more testing is being done. Major results of the work are presented in this paper. The reference V-4Cr-4Ti had the most attractive combination of the mechanical and physical properties that are prerequisite for first-wall and blanket structures: good thermal creep, good tensile strength/ductility, high impact energy, excellent resistance to swelling, and very low ductile-brittle transition temperature before and after irradiation. The alloy was highly resistant to irradiation-induced embrittlement in Li at 420-600 C, and the effects of dynamically charged He on swelling and mechanical properties were insignificant. However, several important issues remain unresolved: welding, low-temperature irradiation, He effect at high dose and high He concentration, irradiation creep, and irradiation performance in air or He. Initial results of investigation of some of these issues are also given.

  16. Perspectives on radiation effects in nickel-base alloys for applications in advanced reactors

    NASA Astrophysics Data System (ADS)

    Rowcliffe, A. F.; Mansur, L. K.; Hoelzer, D. T.; Nanstad, R. K.

    2009-07-01

    Because of their superior high temperature strength and corrosion properties, a set of Ni-base alloys has been proposed for various in-core applications in Gen IV reactor systems. However, irradiation-performance data for these alloys is either limited or non-existent. A review is presented of the irradiation-performance of a group of Ni-base alloys based upon data from fast breeder reactor programs conducted in the 1975-1985 timeframe with emphasis on the mechanisms involved in the loss of high temperature ductility and the breakdown in swelling resistance with increasing neutron dose. The implications of these data for the performance of the Gen IV Ni-base alloys are discussed and possible pathways to mitigate the effects of irradiation on alloy performance are outlined. A radical approach to designing radiation damage-resistant Ni alloys based upon recent advances in mechanical alloying is also described.

  17. Quantitative analytical electron microscopy of multiphase alloys.

    PubMed

    Prybylowski, J; Ballinger, R; Elliott, C

    1989-02-01

    In this paper, we present a technique for analysis of composition gradients, using an analytical electron microscope, within the primary phase of a two-phase alloy for the case where the second-phase particle size is similar to the size of the irradiated volume. If the composition difference between the two phases is large, the detected compositional fluctuations associated with varying phase fractions may mask any underlying composition gradient of the primary phase. The analysis technique was used to determine grain boundary chromium concentration gradients in a nickel-base superalloy, alloy X-750. The technique may also be of use in other alloy systems. PMID:2709131

  18. METHOD OF DISSOLVING REFRACTORY ALLOYS

    DOEpatents

    Helton, D.M.; Savolainen, J.K.

    1963-04-23

    This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

  19. Modelling Thermodynamics of Alloys for Fusion Application

    SciTech Connect

    Caro, A; Sadigh, B; Turchi, P A; Caro, M; Lopasso, E; Crowson, D

    2006-01-26

    This research has two main objectives: (1) On one side is the development of computational tools to evaluate alloy properties, using the information contained in thermodynamic functions to improve the ability of classic potentials to account for complex alloy behavior. (2) On the other hand, to apply the tools so developed to predict properties of alloys under irradiation. Atomistic simulations of alloys at the empirical level face the challenge of correctly modeling basic thermodynamic properties. In this work we develop a methodology to generalize many-body classic potentials to incorporate complex formation energy curves. Application to Fe-Cr allows us to predict the implications of the ab initio results of formation energy on the phase diagram of this alloy.

  20. PLUTONIUM ALLOYS

    DOEpatents

    Chynoweth, W.

    1959-06-16

    The preparation of low-melting-point plutonium alloys is described. In a MgO crucible Pu is placed on top of the lighter alloying metal (Fe, Co, or Ni) and the temperature raised to 1000 or 1200 deg C. Upon cooling, the alloy slug is broke out of the crucible. With 14 at. % Ni the m.p. is 465 deg C; with 9.5 at. % Fe the m.p. is 410 deg C; and with 12.0 at. % Co the m.p. is 405 deg C. (T.R.H.) l6262 l6263 ((((((((Abstract unscannable))))))))

  1. The effect of fusion-relevant helium levels on the mechanical properties of isotopically tailored ferritic alloys

    SciTech Connect

    Hankin, G.L.; Hamilton, M.L.; Gelles, D.S.

    1997-04-01

    The yield and maximum strengths of an irradiated series of isotopically tailored ferritic alloys were evaluated using the shear punch test. The composition of three of the alloys was Fe-12Cr-1.5Ni. Different balances of nickel isotopes were used in each alloy in order to produce different helium levels. A fourth alloy, which contained no nickel, was also irradiated. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys, and as expected, the strength of the alloys decreased with increasing irradiation temperature. Helium itself, up to 75 appm over 7 dpa appears to have little effect on the mechanical properties of the alloys.

  2. BRAZING ALLOYS

    DOEpatents

    Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

    1962-02-20

    A brazing alloy is described which, in the molten state, is characterized by excellent wettability and flowability and is capable of forming a corrosion-resistant brazed joint. At least one component of said joint is graphite and the other component is a corrosion-resistant refractory metal. The brazing alloy consists essentially of 40 to 90 wt % of gold, 5 to 35 wt% of nickel, and 1 to 45 wt% of tantalum. (AEC)

  3. COATED ALLOYS

    DOEpatents

    Harman, C.G.; O'Bannon, L.S.

    1958-07-15

    A coating is described for iron group metals and alloys, that is particularly suitable for use with nickel containing alloys. The coating is glassy in nature and consists of a mixture containing an alkali metal oxide, strontium oxide, and silicon oxide. When the glass coated nickel base metal is"fired'' at less than the melting point of the coating, it appears the nlckel diffuses into the vitreous coating, thus providing a closely adherent and protective cladding.

  4. High Accuracy U-235 Enrichment Verification Station for Low Enriched Uranium Alloys

    SciTech Connect

    Lillard, C. R.; Hayward, J. P.; Williamson, M. R.

    2012-06-07

    The Y-12 National Security Complex is playing a role in the U.S. High Performance Research Reactor (USHPRR) Conversion program sponsored by the U.S. National Nuclear Security Administration's Office of Global Threat Reduction. The USHPRR program has a goal of converting remaining U.S. reactors that continue to use highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The USHPRR program is currently developing a LEU Uranium-Molybdenum (U-Mo) monolithic fuel for use in the U.S. high performance research reactors.Y-12 is supporting both the fuel development and fuel fabrication efforts by fabricating low enriched U-Mo foils from its own source material for irradiation experiments and for optimizing the fabrication process in support of scaling up the process to a commercial production scale. Once the new fuel is qualified, Y-12 will produce and ship U-Mo coupons with verified 19.75% +0.2% - 0.3% U-235 enrichment to be fabricated into fuel elements for the USHPRRs. Considering this small enrichment tolerance and the transition into HEU being set strictly at 20% U-235, a characterization system with a measurement uncertainty of less than or equal to 0.1% in enrichment is desired to support customer requirements and minimize production costs. Typical uncertainty for most available characterization systems today is approximately 1-5%; therefore, a specialized system must be developed which results in a reduced measurement uncertainty. A potential system using a High-Purity Germanium (HPGe) detector has been procured, and tests have been conducted to verify its capabilities with regards to the requirements. Using four U-Mo enrichment standards fabricated with complete isotopic and chemical characterization, infinite thickness and peak-ratio enrichment measurement methods have been considered for use. As a result of inhomogeneity within the U-Mo samples, FRAM, an isotopic analysis software, has been selected for initial testing. A systematic approach

  5. Comparison of properties and microstructures of Trefimetaux and Hycon 3HP{trademark} after neutron irradiation

    SciTech Connect

    Edwards, D.J.; Singh, B.N.; Toft, P.; Eldrup, M.

    1998-09-01

    The precipitation strengthened CuNiBe alloys are among three candidate copper alloys being evaluated for application in the first wall, divertor, and limiter components of ITER. Generally, CuNiBe alloys have higher strength but poorer conductivity compared to CuCrZr and CuAl{sub 2}O{sub 3} alloys. Brush-Wellman Inc. has manufactured an improved version of their Hycon CuNiBe alloy that has higher conductivity while maintaining a reasonable level strength. It is of interest, therefore, to investigate the effect of radiation on the physical and mechanical properties of this alloy. In the present work the authors have investigated the physical and mechanical properties of the Hycon 3HP{trademark} alloy both before and after neutron irradiation and have compared its microstructure and properties with the European CuNiBe candidate alloy manufactured by Trefirmetaux. Tensile specimens of both alloys were irradiated in the DR-3 reactor at Risoe to displacement dose levels up to 0.3 dpa at 100, 250 and 350 C. Both alloys were tensile tested in the unirradiated and irradiated conditions at 100, 250 and 350 C. Both pre- and post-irradiation microstructures of the alloys were investigated in detail using transmission electron microscopy. Fracture surfaces were examined under a scanning electron microscope. Electrical resistivity measurements were made on tensile specimens before and after irradiation; all measurements were made at 23 C. At this point it seems unlikely that CuNiBe alloys can be recommended for applications in neutron environments where the irradiation temperature exceeds 200 C. Applications at temperatures below 200 C might be plausible, but only after careful experiments have determined the dose dependence of the mechanical properties and the effect of sudden temperature excursions on the material to establish the limits on the use of the alloy.

  6. Research and development on vanadium alloys for fusion applications

    SciTech Connect

    Zinkle, S.J.; Rowcliffe, A.F.; Matsui, H.; Abe, K.; Smith, D.L.; Osch, E. van; Kazakov, V.A.

    1998-03-01

    The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

  7. Radiation-induced segregation in alloy X-750

    SciTech Connect

    Kenik, E.A.

    1996-12-31

    Microstructural and microchemical evolution of an Alloy X-750 heat under neutron irradiation was studied in order to understand the origin of irradiation-assisted stress corrosion cracking. Both clustering of point defects and radiation-induced segregation at interfaces were observed. Although no significant changes in the precipitate structure were observed, boundaries exhibited additional depletion of Cr and Fe and enrichment of Ni.

  8. Interaction UMo fuel with Fe and FeCr

    NASA Astrophysics Data System (ADS)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Uranium-molybdenum alloys are perspective nuclear fuel for fast reactors. In this work, a study was conducted of the interaction of uranium-molybdenum alloy with iron and chromium steel at an elevated temperature of 750 °C for 5 hours. It was found that the constant rate of the interaction layer growth for diffusion couple UMo/FeCr is about 5.4-10-12 m2/s at 750 °C. The phase composition of the interaction layers for the both diffusion couples was determined. The interaction comes along the grain boundaries, there are not interacts UMo alloy grain in the structure of the diffusion zone.

  9. Tissue irradiator

    DOEpatents

    Hungate, F.P.; Riemath, W.F.; Bunnell, L.R.

    1975-12-16

    A tissue irradiator is provided for the in-vivo irradiation of body tissue. The irradiator comprises a radiation source material contained and completely encapsulated within vitreous carbon. An embodiment for use as an in- vivo blood irradiator comprises a cylindrical body having an axial bore therethrough. A radioisotope is contained within a first portion of vitreous carbon cylindrically surrounding the axial bore, and a containment portion of vitreous carbon surrounds the radioisotope containing portion, the two portions of vitreous carbon being integrally formed as a single unit. Connecting means are provided at each end of the cylindrical body to permit connections to blood- carrying vessels and to provide for passage of blood through the bore. In a preferred embodiment, the radioisotope is thulium-170 which is present in the irradiator in the form of thulium oxide. A method of producing the preferred blood irradiator is also provided, whereby nonradioactive thulium-169 is dispersed within a polyfurfuryl alcohol resin which is carbonized and fired to form the integral vitreous carbon body and the device is activated by neutron bombardment of the thulium-169 to produce the beta-emitting thulium-170.

  10. Survey of Radiation Effects in Titanium Alloys

    SciTech Connect

    Mansur, Louis K

    2008-08-01

    Information on radiation effects in titanium alloys has been reviewed. Only sparse experimental data from fission reactor and charged particle irradiations is available, none of which is directly applicable to the SNS. Within this limited data it is found that although mechanical properties are substantially degraded, several Ti alloys may retain acceptable properties to low or moderate doses. Therefore, it is recommended that titanium alloys be examined further for application to the SNS target. Since information directly relevant to the SNS mercury target environment and irradiation conditions is not available, it is recommended that ORNL generate the necessary experimental data using a graded approach. The first testing would be for cavitation erosion resistance using two different test devices. If the material performs acceptably the next tests should be for long term mercury compatibility testing of the most promising alloys. Irradiation tests to anticipated SNS displacement doses followed by mechanical property measurements would be the last stage in determining whether the alloys should be considered for service in the SNS target module.

  11. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  12. Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems

    SciTech Connect

    Stephen M. Bruemmer; Peter L. Andersen; Gary Was

    2002-12-27

    The discovery of a damage-resistant alloy based on Hf solute additions to a low-carbon 316SS is the highlight of the Phase II research. This damage resistance is supported by characterization of radiation-induced microstructures and microchemistries along with measurements of environmental cracking. The addition of Hf to a low-carbon 316SS reduced the detrimental impact of radiation by changing the distribution of Hf. Pt additions reduced the impact of radiation on grain boundary segregation but did not alter its effect on microstructural damage development or cracking. Because cracking susceptibility is associated with several material characteristics, separate effect experiments exploring strength effects using non-irradiated stainless steels were conducted. These crack growth tests suggest that irradiation strength by itself can promote environmental cracking. The second concept for developing damage resistant alloys is the use of metastable precipitates to stabilize the microstructure during irradiation. Three alloys have been tailored for evaluation of precipitate stability influences on damage evolution. The first alloy is a Ni-base alloy (alloy 718) that has been characterized at low neutron irradiation doses but has not been characterized at high irradiation doses. The other two alloys are Fe-base alloys (PH 17-7 and PH 17-4) that have similar precipitate structures as alloy 718 but is more practical in nuclear structures because of the lower Ni content and hence lesser transmutation to He.

  13. First principle-based AKMC modelling of the formation and medium-term evolution of point defect and solute-rich clusters in a neutron irradiated complex Fe-CuMnNiSiP alloy representative of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Ngayam-Happy, R.; Becquart, C. S.; Domain, C.

    2013-09-01

    The formation and medium-term evolution of point defect and solute-rich clusters under neutron irradiation have been modelled in a complex Fe-CuMnNiSiP alloy representative of RPV steels, by means of first principle-based atomistic kinetic Monte Carlo simulations. The results obtained reproduce most features observed in available experimental studies, highlighting the very good agreement between both series. According to simulation, solute-rich clusters form and develop via an induced segregation mechanism on either the vacancy or interstitial clusters, and these point defect clusters are efficiently generated only in cascade debris and not Frenkel pair flux. The results have revealed the existence of two distinct populations of clusters with different characteristic features. Solute-rich clusters in the first group are bound essentially to interstitial clusters and they are enriched in Mn mostly, but also Ni to a lesser extent. Over the low dose regime, their density increases in the alloy as a result of the accumulation of highly stable interstitial clusters. In the second group, the solute-rich clusters are merged with vacancy clusters, and they contain mostly Cu and Si, but also substantial amount of Mn and Ni. The formation of a sub-population of pure solute clusters has been observed, which results from annihilation of the low stable vacancy clusters on sinks. The results indicate finally that the Mn content in clusters is up to 50%, Cu, Si, and Ni sharing the other half in more or less equivalent amounts. This composition has not demonstrated any noticeable modification with increasing dose over irradiation.

  14. Irradiation subassembly

    DOEpatents

    Seim, O.S.; Filewicz, E.C.; Hutter, E.

    1973-10-23

    An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

  15. [Radioactivity of phosphorus implanted TiNi alloy].

    PubMed

    Zhao, Xingke; Cai, Wei; Zhao, Liancheng

    2003-09-01

    Exposed to neutron flow, the phosphorus implanted TiNi alloy gets radioactive. This radioactive material is used in vascular stent for prevention and cure of restenosis. Phosphorus implantation is carried out in a plasma immerged ion implantation system, and the dose of phosphorus implantation is in the range of 2-10 x 10(17) cm-2. After ion implantation, the alloy is exposed to the slow neutron flow in a nuclear reactor, the dose of the slow neutron is 1.39-5.88 x 10(19) n/cm2. The radioactivity of the TiNi alloy was measured by liquid scintillation spectrometry and radio-chromic-film dosimetry. The result shows that whether the phosphorus is implanted or not, the TiNi alloy comes to be radioactive after exposure to neutron flow. Just after neutron irradiation, the radiation dose of phosphorus implanted TiNi alloy is about one hundred times higher than that of un-phosphorus implanted TiNi alloy. The radiation difference between phosphorus and un-phosphorus implanted alloy decreases as time elapses. Within three months after neutron irradiation, the average half-decay period of phosphorus implanted TiNi alloy is about 62 days. The radiation ray penetration of phosphorus implanted TiNi alloy is deeper than that of pure 32P; this is of benefit to making radiation uniformity between stent struts and reducing radiation grads beyond the edge of stent.

  16. The correlation between swelling and radiation-induced segregation in iron-chromium-nickel alloys.

    SciTech Connect

    Allen, T. R.; Busby, J. T.; Kenik, E. A.; Was, G. S.

    1998-03-05

    The magnitudes of both void swelling and radiation-induced segregation (RIS) in iron-chromium-nickel alloys are dependent on bulk alloy composition. Because the diffusivity of nickel via the vacancy flux is slow relative to chromium, nickel enriches and chromium depletes at void surfaces during irradiation. This local composition change reduces the subsequent vacancy flux to the void, thereby reducing void swelling. In this work, the resistance to swelling from major element segregation is estimated using diffusivities derived from grain boundary segregation measurements in irradiated iron-chromium-nickel alloys. The resistance to void swelling in iron- and nickel-base alloys correlates with the segregation and both are functions of bulk alloy composition. Alloys that display the greatest amount of nickel enrichment and chromium depletion are found to be most resistant to void swelling, as predicted. Additionally, swelling is shown to be greater in alloys in which the RIS profiles are slow to develop.

  17. Irradiated foods

    MedlinePlus

    ... it reduces the risk of food poisoning . Food irradiation is used in many countries. It was first approved in the U.S. to prevent sprouts on white potatoes, and to control insects on wheat and in certain spices and seasonings.

  18. Fatigue behavior of Type 316 stainless steel following neutron irradiation inducing helium

    SciTech Connect

    Grossbeck, M.L.; Liu, K.C.

    1980-01-01

    Since a tokamak reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied.

  19. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

    NASA Astrophysics Data System (ADS)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

  20. Alloy softening in binary molybdenum alloys

    NASA Technical Reports Server (NTRS)

    Stephens, J. R.; Witzke, W. R.

    1972-01-01

    An investigation was conducted to determine the effects of alloy additions of Hf, Ta, W, Re, Os, Ir, and Pt on the hardness of Mo. Special emphasis was placed on alloy softening in these binary Mo alloys. Results showed that alloy softening was produced by those elements having an excess of s+d electrons compared to Mo, while those elements having an equal number or fewer s+d electrons than Mo failed to produce alloy softening. Alloy softening and hardening can be correlated with the difference in number of s+d electrons of the solute element and Mo.

  1. Small-scale characterisation of irradiated nuclear materials: Part I - Microstructure

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; London, A.; Xu, A.; Armstrong, D. E. J.; Roberts, S. G.

    2015-07-01

    The behaviour of nanometre-scale precipitates in oxide dispersion strengthened (ODS) ferritic alloys and tungsten-rhenium alloys for nuclear applications has been examined by atom probe tomography (APT). Low Re content tungsten alloys showed no evidence of Re clustering following self-ion irradiation whereas the 25 at.% Re resulted in cluster formation. The size and composition of clusters varied depending on the material form during irradiation (pre-sharpened needle or bulk). These results highlight the care that must be taken in interpreting data from ion irradiated pre-sharpened needles due to the presence of free surfaces. Self-ion irradiation of the ODS ferritic alloy resulted in a change in the composition of the clusters, indicating a transition from a near-stoichiometric Y2Ti2O7 composition towards a Ti2YO5.

  2. Small-scale characterisation of irradiated nuclear materials: Part I – Microstructure

    DOE PAGES

    Edmondson, P. D.; London, A.; Xu, A.; Armstrong, D. E. J.; Roberts, S. G.

    2014-11-26

    The behaviour of nanometre-scale precipitates in oxide dispersion strengthened (ODS) ferritic alloys and tungsten-rhenium alloys for nuclear applications has been examined by atom probe tomography (APT). Low Re content tungsten alloys showed no evidence of Re clustering following self-ion irradiation whereas the 25 at.% Re resulted in cluster formation. The size and composition of clusters varied depending on the material form during irradiation (pre-sharpened needle or bulk). Lastly, these results highlight the care that must be taken in interpreting data from ion irradiated pre-sharpened needles due to the presence of free surfaces. Self-ion irradiation of the ODS ferritic alloy resultedmore » in a change in the composition of the clusters, indicating a transition from a near-stoichiometric Y2Ti2O7 composition towards a Ti2YO5.« less

  3. Small-scale characterisation of irradiated nuclear materials: Part I – Microstructure

    SciTech Connect

    Edmondson, P. D.; London, A.; Xu, A.; Armstrong, D. E. J.; Roberts, S. G.

    2014-11-26

    The behaviour of nanometre-scale precipitates in oxide dispersion strengthened (ODS) ferritic alloys and tungsten-rhenium alloys for nuclear applications has been examined by atom probe tomography (APT). Low Re content tungsten alloys showed no evidence of Re clustering following self-ion irradiation whereas the 25 at.% Re resulted in cluster formation. The size and composition of clusters varied depending on the material form during irradiation (pre-sharpened needle or bulk). Lastly, these results highlight the care that must be taken in interpreting data from ion irradiated pre-sharpened needles due to the presence of free surfaces. Self-ion irradiation of the ODS ferritic alloy resulted in a change in the composition of the clusters, indicating a transition from a near-stoichiometric Y2Ti2O7 composition towards a Ti2YO5.

  4. Modeling of microstructure evolution in austenitic stainless steels irradiated under light water reactor condition

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.; Stoller, R. E.

    2001-10-01

    A model for microstructure development in austenitic alloys under light water reactor irradiation conditions is described. The model is derived from the model developed by Stoller and Odette to describe microstructural evolution under fast neutron or fusion reactor irradiation conditions. The model is benchmarked against microstructure measurements in 304 and 316 SS irradiated in a boiling water reactor core using one material-dependent and three irradiation-based parameters. The model is also adapted for proton irradiation at higher dose rate and higher temperature and is calibrated against microstructure measurements for proton irradiation. The model calculations show that for both neutron and proton irradiations, in-cascade interstitial clustering is the driving mechanism for loop nucleation. The loss of interstitial clusters to sinks by interstitial cluster diffusion was found to be an important factor in determining the loop density. The model also explains how proton irradiation can produce an irradiated dislocation microstructure similar to that in neutron irradiation.

  5. Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

    DOE PAGES

    Parish, Chad M.; Field, Kevin G.; Certain, Alicia G.; Wharry, Janelle P.

    2015-04-20

    This paper provides a general overview of advanced scanning transmission electron microscopy (STEM) techniques used for characterization of irradiated BCC Fe-based alloys. Advanced STEM methods provide the high-resolution imaging and chemical analysis necessary to understand the irradiation response of BCC Fe-based alloys. The use of STEM with energy dispersive x-ray spectroscopy (EDX) for measurement of radiation-induced segregation (RIS) is described, with an illustrated example of RIS in proton- and self-ion irradiated T91. Aberration-corrected STEM-EDX for nanocluster/nanoparticle imaging and chemical analysis is also discussed, and examples are provided from ion-irradiated oxide dispersion strengthened (ODS) alloys. In conclusion, STEM techniques for void,more » cavity, and dislocation loop imaging are described, with examples from various BCC Fe-based alloys.« less

  6. Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

    SciTech Connect

    Parish, Chad M.; Field, Kevin G.; Certain, Alicia G.; Wharry, Janelle P.

    2015-04-20

    This paper provides a general overview of advanced scanning transmission electron microscopy (STEM) techniques used for characterization of irradiated BCC Fe-based alloys. Advanced STEM methods provide the high-resolution imaging and chemical analysis necessary to understand the irradiation response of BCC Fe-based alloys. The use of STEM with energy dispersive x-ray spectroscopy (EDX) for measurement of radiation-induced segregation (RIS) is described, with an illustrated example of RIS in proton- and self-ion irradiated T91. Aberration-corrected STEM-EDX for nanocluster/nanoparticle imaging and chemical analysis is also discussed, and examples are provided from ion-irradiated oxide dispersion strengthened (ODS) alloys. In conclusion, STEM techniques for void, cavity, and dislocation loop imaging are described, with examples from various BCC Fe-based alloys.

  7. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    SciTech Connect

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  8. Materials for cold neutron sources: Cryogenic and irradiation effects

    SciTech Connect

    Alexander, D.J.

    1990-01-01

    Materials for the construction of cold neutron sources must satisfy a range of demands. The cryogenic temperature and irradiation create a severe environment. Candidate materials are identified and existing cold sources are briefly surveyed to determine which materials may be used. Aluminum- and magnesium-based alloys are the preferred materials. Existing data for the effects of cryogenic temperature and near-ambient irradiation on the mechanical properties of these alloys are briefly reviewed, and the very limited information on the effects of cryogenic irradiation are outlined. Generating mechanical property data under cold source operating conditions is a daunting prospect. It is clear that the cold source material will be degraded by neutron irradiation, and so the cold source must be designed as a brittle vessel. The continued effective operation of many different cold sources at a number of reactors makes it clear that this can be accomplished. 46 refs., 8 figs., 2 tab.

  9. Modeling microstructure evolution of binary systems subjected to irradiation and mechanical loading

    NASA Astrophysics Data System (ADS)

    Kharchenko, Dmitrii O.; Shchokotova, Olga M.; Lysenko, Irina O.; Kharchenko, Vasyl O.

    2015-07-01

    We study a change in mechanical properties of binary systems subjected to irradiation influence described by ballistic flux of atomic mixing having regular and stochastic contributions. By using numerical modeling based on the phase field approach we study dynamics of deformation fields in a previously irradiated system and in the binary system deformed during irradiation. An influence of both deterministic and stochastic components of ballistic flux onto both yield strength and ultimate strength is studied. We have found that degradation of mechanical properties relates to the formation of percolating clusters of shear bands. Considering a hardening coefficient we analyze stages of plastic deformation of both initially irradiated alloy and alloy subjected to sustained irradiation. Stability of binary alloy under mechanical loading in the form of shear strain with a constant rate and cyclic deformation is discussed.

  10. Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

    SciTech Connect

    Zinkle, S.J.

    1998-03-01

    The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of {approximately}5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule.

  11. Austenitic alloy and reactor components made thereof

    DOEpatents

    Bates, John F.; Brager, Howard R.; Korenko, Michael K.

    1986-01-01

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  12. Metal alloy identifier

    DOEpatents

    Riley, William D.; Brown, Jr., Robert D.

    1987-01-01

    To identify the composition of a metal alloy, sparks generated from the alloy are optically observed and spectrographically analyzed. The spectrographic data, in the form of a full-spectrum plot of intensity versus wavelength, provide the "signature" of the metal alloy. This signature can be compared with similar plots for alloys of known composition to establish the unknown composition by a positive match with a known alloy. An alternative method is to form intensity ratios for pairs of predetermined wavelengths within the observed spectrum and to then compare the values of such ratios with similar values for known alloy compositions, thereby to positively identify the unknown alloy composition.

  13. Electrochemical behavior of nano and femtosecond laser textured titanium alloy for implant surface modification.

    PubMed

    Jeong, Yong-Hoon; Kim, Won-Gi; Choe, Han-Cheol

    2011-02-01

    In this study, the electrochemical behavior of nano and femtosecond laser textured titanium alloy for implant surface modification has been researched using the potentiostat equipment. Cp-Ti and Ti-6Al-4V alloy, located on X-Y motorized stage, were irradiated using femtosecond laser. The corrosion properties were examined by a potentiodynamic and AC impedance test.

  14. Corrosion Resistance of Various High Chromium Alloys in Simulated Chemical Processing Nuclear Plant Waste Solutions

    SciTech Connect

    Anderson, P.A.; Agarwal, D.C.

    1997-12-31

    High chromium nickel alloys were tested at the Idaho Chemical Processing Plant (ICPP) to determine their corrosion performance in the high temperature aggressive chemical environments of liquid waste evaporators used in the chemical reprocessing of irradiated nuclear fuels. The results of these tests, which included a variety of base metal alloys I weld filler material combinations, are presented and discussed.

  15. Irradiation effects in ferritic steels

    NASA Astrophysics Data System (ADS)

    Lechtenberg, Thomas

    1985-08-01

    Since 1979 the Alloy Development for Irradiation Performance (ADIP) task funded by the US Department of Energy has been studying the 2-12Cr class of ferritic steels to establish the feasibility of using them in fusion reactor first wall/breeding blanket (FW/B) applications. The advantages of ferritic steels include superior swelling resistance, low thermal stresses compared to austenitic stainless steels, attractive mechanical properties up to 600°C. and service histories exceeding 100 000 h. These steels are commonly used in a range of microstructural conditions which include ferritic, martensitic. tempered martensitic, bainitic etc. Throughout this paper where the term "ferritic" is used it should be taken to mean any of these microstructures. The ADIP task is studying several candidate alloy systems including 12Cr-1MoWV (HT-9), modified 9Cr-1MoVNb, and dual-phased steels such as EM-12 and 2 {1}/{4}Cr-Mo. These materials are ferromagnetic (FM), body centered cubic (bcc), and contain chromium additions between 2 and 12 wt% and molybdenum additions usually below 2%. The perceived issues associated with the application of this class of steel to fusion reactors are the increase in the ductile-brittle transition temperature (DBTT) with neutron damage, the compatibility of these steels with liquid metals and solid breeding materials, and their weldability. The ferromagnetic character of these steels can also be important in reactor design. It is the purpose of this paper to review the current understanding of these bcc steels and the effects of irradiation. The major points of discussion will be irradiation-induced or -enhanced dimensional changes such as swelling and creep, mechanical properties such as tensile strength and various measurements of toughness, and activation by neutron interactions with structural materials.

  16. Effect of helium on tensile properties of vanadium alloys

    SciTech Connect

    Chung, H.M.; Billone, M.C.; Smith, D.L.

    1997-08-01

    Tensile properties of V-4Cr-4Ti (Heat BL-47), 3Ti-1Si (BL-45), and V-5Ti (BL-46) alloys after irradiation in a conventional irradiation experiment and in the Dynamic Helium Charging Experiment (DHCE) were reported previously. This paper presents revised tensile properties of these alloys, with a focus on the effects of dynamically generated helium of ductility and work-hardening capability at <500{degrees}C. After conventional irradiation (negligible helium generation) at {approx}427{degrees}C, a 30-kg heat of V-4Cr-4Ti (BL-47) exhibited very low uniform elongation, manifesting a strong susceptibility to loss of work-hardening capability. In contrast, a 15-kg heat of V-3Ti-1Si (BL -45) exhibited relatively high uniform elongation ({approx}4%) during conventional irradiation at {approx}427{degrees}C, showing that the heat is resistant to loss of work-hardening capability.

  17. Refractory alloy technology for space nuclear power applications

    SciTech Connect

    Cooper, R.H. Jr.; Hoffman, E.E.

    1984-01-01

    Purpose of this symposium is twofold: (1) to review and document the status of refractory alloy technology for structural and fuel-cladding applications in space nuclear power systems, and (2) to identify and document the refractory alloy research and development needs for the SP-100 Program in both the short and the long term. In this symposium, an effort was made to recapture the space reactor refractory alloy technology that was cut off in midstream around 1973 when the national space nuclear reactor program began in the early 1960s, was terminated. The six technical areas covered in the program are compatibility, processing and production, welding and component fabrication, mechanical and physical properties, effects of irradiation, and machinability. The refractory alloys considered are niobium, molybdenum, tantalum, and tungsten. Thirteen of the 14 pages have been abstracted separately. The remaining paper summarizes key needs for further R and D on refractory alloys. (DLC)

  18. BWR control-rod cobalt-alloy replacement. Final report

    SciTech Connect

    Aldred, P.

    1982-03-01

    Cobalt base pin and roller alloys in BWR Control Rods are a source for the Co-60 isotope which contributes to radiation buildup in the BWR core, the recirculation piping system and the spent fuel pool. It thereby influences personnel radiation exposure during BWR plant maintenance. The program objectives were (a) to identify non-cobalt alloys which could potentially replace the cobalt alloys, (b) evaluate the alloys by testing to qualify them for in-reactor surveillance testing, and (c) to initiate reactor tests at 2 BWRs. Wear resistance, an important requirement for pins and rollers, was measured in a simulated BWR environment (excluding irradiation). Prototypic wear tests were emphasized and a prototype control rod drive test facility was used to evaluate several pin and roller alloy combinations during prototype control rod operations.

  19. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    SciTech Connect

    Grossbeck, M.L.; Gibson, L.T.; Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  20. Ion-irradiation-assisted phase selection in single crystalline Fe7Pd3 ferromagnetic shape memory alloy thin films: from fcc to bcc along the Nishiyama-Wassermann path.

    PubMed

    Arabi-Hashemi, A; Mayr, S G

    2012-11-01

    When processing Fe-Pd ferromagnetic shape memory thin films, selection of the desired phases and their transformation temperatures constitutes one of the largest challenges from an application point of view. In the present contribution we demonstrate that irradiation with 1.8 MeV Kr(+) ions is the method of choice to achieve this goal: Single crystalline Fe(7)Pd(3) thin films that are grown with molecular beam epitaxy on MgO (001) substrates and subsequently irradiated with ions reveal a phase transformation along the whole phase transformation path ranging from fcc austenite to bcc martensite. While for 10(14) ions/cm(2) a fcc-fct phase transformation is observed, increasing the fluence to 5 × 10(14) ions/cm(2) and 5 × 10(15) ions/cm(2) leads to a phase transformation to the bcc phase. Pole figure measurements reveal an orientation relationship for the fcc-bcc phase transformation according to Nishiyama and Wassermann.

  1. Irradiation performance of nitride fuels

    SciTech Connect

    Matthews, R.B.

    1993-01-01

    The properties and advantages of nitride fuels are well documented in the literature. Basically the high thermal conductivity and uranium density of nitride fuels permit high power density, good breeding ratios, low reactivity swings, and large diameter pins compared to oxides. Nitrides are compatible with cladding alloys and liquid metal coolants, thereby reducing fuel/cladding chemical interactions and permitting the use of sodium-bonded pins and the operation of breached pins. Recent analyses done under similar operating conditions show that - compared to metal - fuels mixed nitrides operate at lower temperatures, produce less cladding strain, have greater margins to failure, result in lower transient temperatures, and have lower sodium void reactivity. Uranium nitride fuel pellet fabrication processes were demonstrated during the SP-100 program, and irradiated nitride fuels can be reprocessed by the PUREX process. Irradiation performance data suggest that nitrides have low fission gas release and swelling rates thereby permitting favorable pin designs and long lifetime. The objective of this report is to summarize the available nitride irradiation performance data base and to recommend optimum nitride characteristics for use in advanced liquid metal reactors.

  2. Deformation in metals after low temperature irradiation: Part II - Irradiation hardening, strain hardening, and stress ratios

    SciTech Connect

    Byun, Thak Sang; Li, Meimei

    2008-03-01

    Effects of irradiation at temperatures 200oC on tensile stress parameters are analyzed for dozens of bcc, fcc, and hcp pure metals and alloys, focusing on irradiation hardening, strain hardening, and relationships between the true stress parameters. Similar irradiation-hardening rates are observed for all the metals irrespective of crystal type; typically, the irradiation-hardening rates are large, in the range 100 - 1000 GPa/dpa, at the lowest dose of <0.0001 dpa and decrease with dose to a few tens of MPa/dpa or less at about 10 dpa. However, average irradiation-hardening rates over the dose range of 0 dpa − (the dose to plastic instability at yield) are considerably lower for stainless steels due to their high uniform ductility. It is shown that whereas low temperature irradiation increases the yield stress, it does not significantly change the strain-hardening rate of metallic materials; it decreases the fracture stress only when non-ductile failure occurs. Such dose independence in strain hardening behavior results in strong linear relationships between the true stress parameters. Average ratios of plastic instability stress to unirradiated yield stress are about 1.4, 3.9, and 1.3 for bcc metals (and precipitation hardened IN718 alloy), annealed fcc metals (and pure Zr), and Zr-4 alloy, respectively. Ratios of fracture stress to plastic instability stress are calculated to be 2.2, 1.7, and 2.1, respectively. Comparison of these values confirms that the annealed fcc metals and other soft metals have larger uniform ductility but smaller necking ductility when compared to other materials.

  3. The role of electro-explosion alloying with titanium diboride and treatment with pulsed electron beam in the surface modification of VT6 alloy

    SciTech Connect

    Konovalov, Sergey Gromov, Victor Kobzareva, Tatyana; Semina, Olga; Bataev, Vladimir; Ivanov, Yurii

    2015-10-27

    The paper presents the results of the investigation of VT6 titanium alloy subjected to electro-explosion alloying with TiB{sub 2} and irradiation with pulsed electron beam. It was established that electro-explosion alloying resulted in a high level of roughness of the surface layer with high adhesion of the modified layer and matrix. Further irradiation of the material with electron beam resulted in the smoothing of the surface of alloying and formation of a porous structure with various scale levels in the surface layer. It was also established that the energetic exposure causes the formation of a gradient structure with a changing elemental composition along the direction from the surface of alloying.

  4. Development of High-Temperature Ferritic Alloys and Performance Prediction Methods for Advanced Fission Energy Systems

    SciTech Connect

    G. RObert Odette; Takuya Yamamoto

    2009-08-14

    Reports the results of a comprehensive development and analysis of a database on irradiation hardening and embrittlement of tempered martensitic steels (TMS). Alloy specific quantitative semi-empirical models were derived for the dpa dose, irradiation temperature (ti) and test (Tt) temperature of yield stress hardening (or softening) .

  5. Effects of alloying elements on thermal desorption of helium in Ni alloys

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Cao, X. Z.; Sato, K.; Yoshiie, T.

    2012-12-01

    It is well known that the minor elements Si and Sn can suppress the formation of voids in Ni alloys. In the present study, to investigate the effects of Si and Sn on the retention of helium in Ni alloys, Ni, Ni-Si, and Ni-Sn alloys were irradiated by 5 keV He ions at 723 K. Thermal desorption spectroscopy (TDS) was performed at up to 1520 K, and microstructural observations were carried out to identify the helium trapping sites during the TDS analysis. Two peaks, at 1350 and 1457 K, appeared in the TDS spectrum of Ni. On the basis of the microstructural observations, the former peak was attributed to the release of trapped helium from small cavities and the latter to its release from large cavities. Small-cavity helium trapping sites were also found in the Ni-Si and Ni-Sn alloys, but no large cavities were observed in these alloys. In addition, it was found that the oversized element Sn could trap He atoms in the Ni-Sn alloy.

  6. Use of Irradiated Foods

    NASA Technical Reports Server (NTRS)

    Brynjolfsson, A.

    1985-01-01

    The safety of irradiated foods is reviewed. Guidelines and regulations for processing irradiated foods are considered. The radiolytic products formed in food when it is irradiated and its wholesomeness is discussed. It is concluded that food irradiation processing is not a panacea for all problems in food processing but when properly used will serve the space station well.

  7. Influence of chemical disorder on energy dissipation and defect evolution in concentrated solid solution alloys

    SciTech Connect

    Zhang, Yanwen; Stocks, George Malcolm; Jin, Ke; Lu, Chenyang; Bei, Hongbin; Sales, Brian C.; Wang, Lumin; Béland, Laurent K.; Stoller, Roger E.; Samolyuk, German D.; Caro, Magdalena; Caro, Alfredo; Weber, William J.

    2015-10-28

    A long-standing objective in materials research is to understand how energy is dissipated in both the electronic and atomic subsystems in irradiated materials, and how related non-equilibrium processes may affect defect dynamics and microstructure evolution. Here we show that alloy complexity in concentrated solid solution alloys having both an increasing number of principal elements and altered concentrations of specific elements can lead to substantial reduction in the electron mean free path and thermal conductivity, which has a significant impact on energy dissipation and consequentially on defect evolution during ion irradiation. Enhanced radiation resistance with increasing complexity from pure nickel to binary and to more complex quaternary solid solutions is observed under ion irradiation up to an average damage level of 1 displacement per atom. Understanding how materials properties can be tailored by alloy complexity and their influence on defect dynamics may pave the way for new principles for the design of radiation tolerant structural alloys.

  8. Irradiation-induced composition patterns in binary solid solutions

    SciTech Connect

    Dubey, Santosh; El-Azab, Anter

    2013-09-28

    A theoretical/computational model for the irradiation-driven compositional instabilities in binary solid solutions has been developed. The model is suitable for investigating the behavior of structural alloys and metallic nuclear fuels in a reactor environment as well as the response of alloy thin films to ion beam irradiation. The model is based on a set of reaction-diffusion equations for the dynamics of vacancies, interstitials, and lattice atoms under irradiation. The dynamics of these species includes the stochastic generation of defects by collision cascades as well as the defect reactions and diffusion. The atomic fluxes in this model are derived based on the transitions of lattice defects. The set of reaction-diffusion equations are stiff, hence a stiffly stable method, also known as the Gear method, has been used to numerically approximate the equations. For the Cu-Au alloy in the solid solution regime, the model results demonstrate the formation of compositional patterns under high-temperature particle irradiation, with Fourier space properties (Fourier spectrum, average wavelength, and wavevector) depending on the cascade damage characteristics, average composition, and irradiation temperature.

  9. Application of Laser Design of Amorphous Feco-Based Alloys for the Formation of Amorphous-Crystalline Composites

    NASA Astrophysics Data System (ADS)

    Permyakova, I. E.; Glezer, A. M.; Ivanov, A. A.; Shelyakov, A. V.

    2016-01-01

    Morphological and fractographic features of change of FeCo-based amorphous alloy surfaces after laser treatment are studied in detail. Regimes of laser treatment that allow various degrees of crystallization of the examined alloys to be obtained, including thin (<1 •m) crystal layers on amorphous alloy surfaces, amorphous-crystalline composites, and completely crystalline alloys are adjusted. The Vickers hardness is estimated in zones of selective laser irradiation. The structure of the examined alloys attendant to the change of their mechanical properties is analyzed.

  10. Turbine Blade Alloy

    NASA Technical Reports Server (NTRS)

    MacKay, Rebecca

    2001-01-01

    The High Speed Research Airfoil Alloy Program developed a fourth-generation alloy with up to an +85 F increase in creep rupture capability over current production airfoil alloys. Since improved strength is typically obtained when the limits of microstructural stability are exceeded slightly, it is not surprising that this alloy has a tendency to exhibit microstructural instabilities after high temperature exposures. This presentation will discuss recent results obtained on coated fourth-generation alloys for subsonic turbine blade applications under the NASA Ultra-Efficient Engine Technology (UEET) Program. Progress made in reducing microstructural instabilities in these alloys will be presented. In addition, plans will be presented for advanced alloy development and for computational modeling, which will aid future alloy development efforts.

  11. Detection of irradiated liquor

    NASA Astrophysics Data System (ADS)

    Shengchu, Qi; Jilan, Wu; Rongyao, Yuan

    D-2,3-butanediol is formed by irradiation processes in irradiated liquors. This radiolytic product is not formed in unirradiated liquors and its presence can therefore be used to identify whether a liquor has been irradiated or not. The relation meso/dl≈1 for 2,3-butanediol and the amount present in irradiated liquors may therefore be used as an indication of the dose used in the irradiation.

  12. Low temperature embrittlement behaviour of different ferritic-martensitic alloys for fusion applications

    NASA Astrophysics Data System (ADS)

    Rieth, M.; Dafferner, B.

    1996-10-01

    In the last few years a lot of different low activation CrWVTa steels have been developed world-wide. Without irradiation some of these alloys show clearly a better low temperature embrittlement behaviour than commercial CrNiMoV(Nb) alloys. Within the MANITU project a study was carried out to compare, prior to the irradiation program, the embrittlement behaviour of different alloys in the unirradiated condition performing instrumented Charpy impact bending tests with sub-size specimens. The low activation materials (LAM) considered were different OPTIFER alloys (Forschungszentrum Karlsruhe), F82H (JAERI), 9Cr2WVTa (ORNL), and GA3X (PNL). The modified commercial 10-11% CrNiMoVNb steels were MANET and OPTIMAR. A meaningful comparison between these alloys could be drawn, since the specimens of all materials were manufactured and tested under the same conditions.

  13. Advances in Development of Vanadium Alloys and MHD Insulator Coatings

    SciTech Connect

    Muroga, Takeo; Chen, J M; Chernov, V M; Fukumoto, K; Hoelzer, David T; Kurtz, Richard; Nagasaka, T; Pint, Bruce A; Satou, M; Suzuki, Akihiro; Watanabe, H

    2007-01-01

    Recent progress in the development of low activation vanadium alloys and MHD insulator coatings for a Li-self cooled blanket is reviewed. Research progress in vanadium alloys is highlighted by technology for fabricating creep tubes, comparison of thermal creep in vacuum and Li, understanding impurity transfer between vanadium alloys and Li and its impact on mechanical properties, behavior of hydrogen and hydrogen isotopes, low dose irradiation effects on weld joints, and exploration for advanced vanadium alloys. Major remaining issues for vanadium alloys are thermal and irradiation creep, helium effects on high-temperature mechanical properties and radiation effects on low-temperature fracture properties. Er2O3 showed good compatibility with Li, and is promising as a MHD insulator coating on vanadium alloys. Significant progress in coating technology for this material has been made. Recent efforts are focused on multi-layer and in-situ coatings. Tests under flowing lithium conditions with a temperature gradient are necessary for quantitative examination of coating performance.

  14. Separation in Binary Alloys

    NASA Technical Reports Server (NTRS)

    Frazier, D. O.; Facemire, B. R.; Kaukler, W. F.; Witherow, W. K.; Fanning, U.

    1986-01-01

    Studies of monotectic alloys and alloy analogs reviewed. Report surveys research on liquid/liquid and solid/liquid separation in binary monotectic alloys. Emphasizes separation processes in low gravity, such as in outer space or in free fall in drop towers. Advances in methods of controlling separation in experiments highlighted.

  15. SUPERCONDUCTING VANADIUM BASE ALLOY

    DOEpatents

    Cleary, H.J.

    1958-10-21

    A new vanadium-base alloy which possesses remarkable superconducting properties is presented. The alloy consists of approximately one atomic percent of palladium, the balance being vanadium. The alloy is stated to be useful in a cryotron in digital computer circuits.

  16. PLUTONIUM-THORIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.

    1959-09-15

    New plutonium-base binary alloys useful as liquid reactor fuel are described. The alloys consist of 50 to 98 at.% thorium with the remainder plutonium. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are easy fabrication, phase stability, and the accompanying advantuge of providing a means for converting Th/sup 232/ into U/sup 233/.

  17. DELTA PHASE PLUTONIUM ALLOYS

    DOEpatents

    Cramer, E.M.; Ellinger, F.H.; Land. C.C.

    1960-03-22

    Delta-phase plutonium alloys were developed suitable for use as reactor fuels. The alloys consist of from 1 to 4 at.% zinc and the balance plutonium. The alloys have good neutronic, corrosion, and fabrication characteristics snd possess good dimensional characteristics throughout an operating temperature range from 300 to 490 deg C.

  18. White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

    SciTech Connect

    Sindelar, R.; Louthan, M.; PNNL, B.

    2015-05-29

    This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

  19. Review of recent irradiation-creep results

    SciTech Connect

    Coghlan, W.A.

    1982-05-01

    Materials deform faster under stress in the presence of irradiation by a process known as irradiation creep. This phenomenon is important to reactor design and has been the subject of a large number of experimental and theoretical investigations. The purpose of this work is to review the recent experimental results to obtain a summary of these results and to determine those research areas that require additional information. The investigations have been classified into four subgroups based on the different experimental methods used. These four are: (1) irradiation creep using stress relaxation methods, (2) creep measurements using pressurized tubes, (3) irradiation creep from constant applied load, and (4) irradiation creep experiments using accelerated particles. The similarity and the differences of the results from these methods are discussed and a summary of important results and suggested areas for research is presented. In brief, the important results relate to the dependence of creep on swelling, temperature, stress state and alloying additions. In each of these areas new results have been presented and new questions have arisen which require further research to answer. 65 references.

  20. Emulation of reactor irradiation damage using ion beams

    SciTech Connect

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  1. Emulation of reactor irradiation damage using ion beams

    DOE PAGES

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  2. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    SciTech Connect

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

  3. Sonocatalytic injury of cancer cells attached on the surface of a nickel-titanium dioxide alloy plate.

    PubMed

    Ninomiya, Kazuaki; Maruyama, Hirotaka; Ogino, Chiaki; Takahashi, Kenji; Shimizu, Nobuaki

    2016-01-01

    The present study demonstrates ultrasound-induced cell injury using a nickel-titanium dioxide (Ni-TiO2) alloy plate as a sonocatalyst and a cell culture surface. Ultrasound irradiation of cell-free Ni-TiO2 alloy plates with 1 MHz ultrasound at 0.5 W/cm(2) for 30s led to an increased generation of hydroxyl (OH) radicals compared to nickel-titanium (Ni-Ti) control alloy plates with and without ultrasound irradiation. When human breast cancer cells (MCF-7 cells) cultured on the Ni-TiO2 alloy plates were irradiated with 1 MHz ultrasound at 0.5 W/cm(2) for 30s and then incubated for 48 h, cell density on the alloy plate was reduced to approximately 50% of the controls on the Ni-Ti alloy plates with and without ultrasound irradiation. These results indicate the injury of MCF-7 cells following sonocatalytic OH radical generation by Ni-TiO2. Further experiments demonstrated cell shrinkage and chromatin condensation after ultrasound irradiation of MCF-7 cells attached on the Ni-TiO2 alloy plates, indicating induction of apoptosis.

  4. High strength alloys

    DOEpatents

    Maziasz, Phillip James [Oak Ridge, TN; Shingledecker, John Paul [Knoxville, TN; Santella, Michael Leonard [Knoxville, TN; Schneibel, Joachim Hugo [Knoxville, TN; Sikka, Vinod Kumar [Oak Ridge, TN; Vinegar, Harold J [Bellaire, TX; John, Randy Carl [Houston, TX; Kim, Dong Sub [Sugar Land, TX

    2010-08-31

    High strength metal alloys are described herein. At least one composition of a metal alloy includes chromium, nickel, copper, manganese, silicon, niobium, tungsten and iron. System, methods, and heaters that include the high strength metal alloys are described herein. At least one heater system may include a canister at least partially made from material containing at least one of the metal alloys. At least one system for heating a subterranean formation may include a tubular that is at least partially made from a material containing at least one of the metal alloys.

  5. High strength alloys

    DOEpatents

    Maziasz, Phillip James; Shingledecker, John Paul; Santella, Michael Leonard; Schneibel, Joachim Hugo; Sikka, Vinod Kumar; Vinegar, Harold J.; John, Randy Carl; Kim, Dong Sub

    2012-06-05

    High strength metal alloys are described herein. At least one composition of a metal alloy includes chromium, nickel, copper, manganese, silicon, niobium, tungsten and iron. System, methods, and heaters that include the high strength metal alloys are described herein. At least one heater system may include a canister at least partially made from material containing at least one of the metal alloys. At least one system for heating a subterranean formation may include a tublar that is at least partially made from a material containing at least one of the metal alloys.

  6. The Irradiation Performance and Microstructural Evolution in 9Cr-2W Steel Under Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Alsagabi, Sultan; Charit, Indrajit; Pasebani, Somayeh

    2016-02-01

    Grade 92 steel (9Cr-2W) is a ferritic-martensitic steel with good mechanical and thermal properties. It is being considered for structural applications in Generation IV reactors. Still, the irradiation performance of this alloy needs more investigation as a result of the limited available data. The irradiation performance investigation of Grade 92 steel would contribute to the understanding of engineering aspects including feasibility of application, economy, and maintenance. In this study, Grade 92 steel was irradiated by iron ion beam to 10, 50, and 100 dpa at 30 and 500 °C. In general, the samples exhibited good radiation damage resistance at these testing parameters. The radiation-induced hardening was higher at 30 °C with higher dislocation density; however, the dislocation density was less pronounced at higher temperature. Moreover, the irradiated samples at 30 °C had defect clusters and their density increased at higher doses. On the other hand, dislocation loops were found in the irradiated sample at 50 dpa and 500 °C. Further, the irradiated samples did not show any bubble or void.

  7. Creep Resistant Zinc Alloy

    SciTech Connect

    Frank E. Goodwin

    2002-12-31

    This report covers the development of Hot Chamber Die Castable Zinc Alloys with High Creep Strengths. This project commenced in 2000, with the primary objective of developing a hot chamber zinc die-casting alloy, capable of satisfactory service at 140 C. The core objectives of the development program were to: (1) fill in missing alloy data areas and develop a more complete empirical model of the influence of alloy composition on creep strength and other selected properties, and (2) based on the results from this model, examine promising alloy composition areas, for further development and for meeting the property combination targets, with the view to designing an optimized alloy composition. The target properties identified by ILZRO for an improved creep resistant zinc die-casting alloy were identified as follows: (1) temperature capability of 1470 C; (2) creep stress of 31 MPa (4500 psi); (3) exposure time of 1000 hours; and (4) maximum creep elongation under these conditions of 1%. The project was broadly divided into three tasks: (1) Task 1--General and Modeling, covering Experimental design of a first batch of alloys, alloy preparation and characterization. (2) Task 2--Refinement and Optimization, covering Experimental design of a second batch of alloys. (3) Task 3--Creep Testing and Technology transfer, covering the finalization of testing and the transfer of technology to the Zinc industry should have at least one improved alloy result from this work.

  8. Tensile and toughness assessment of the procured advanced alloys

    SciTech Connect

    Tan, Lizhen; Sokolov, Mikhail A.; Hoelzer, David T.; Busby, Jeremy T.

    2015-09-11

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance by 2024 in light water reactor (LWR)-relevant environments

  9. Weldability of High Alloys

    SciTech Connect

    Maroef, I

    2003-01-22

    The purpose of this study was to investigate the effect of silicon and iron on the weldability of HAYNES HR-160{reg_sign} alloy. HR-I60 alloy is a solid solution strengthened Ni-Co-Cr-Si alloy. The alloy is designed to resist corrosion in sulfidizing and other aggressive high temperature environments. Silicon is added ({approx}2.75%) to promote the formation of a protective oxide scale in environments with low oxygen activity. HR-160 alloy has found applications in waste incinerators, calciners, pulp and paper recovery boilers, coal gasification systems, and fluidized bed combustion systems. HR-160 alloy has been successfully used in a wide range of welded applications. However, the alloy can be susceptible to solidification cracking under conditions of severe restraint. A previous study by DuPont, et al. [1] showed that silicon promoted solidification cracking in the commercial alloy. In earlier work conducted at Haynes, and also from published work by DuPont et al., it was recognized that silicon segregates to the terminal liquid, creating low melting point liquid films on solidification grain boundaries. Solidification cracking has been encountered when using the alloy as a weld overlay on steel, and when joining HR-160 plate in a thickness greater than19 millimeters (0.75 inches) with matching filler metal. The effect of silicon on the weldability of HR-160 alloy has been well documented, but the effect of iron is not well understood. Prior experience at Haynes has indicated that iron may be detrimental to the solidification cracking resistance of the alloy. Iron does not segregate to the terminal solidification product in nickel-base alloys, as does silicon [2], but iron may have an indirect or interactive influence on weldability. A set of alloys covering a range of silicon and iron contents was prepared and characterized to better understand the welding metallurgy of HR-160 alloy.

  10. Alloying of aluminum-beryllium alloys

    NASA Astrophysics Data System (ADS)

    Molchanova, L. V.; Ilyushin, V. N.

    2013-01-01

    The existing phase diagrams of Al-Be- X alloys, where X is an alloying element, are analyzed. Element X is noted to poorly dissolve in both aluminum and beryllium. It is shown that the absence of intermetallic compounds in the Al-Be system affects the phase equilibria in an Al-Be- X system. Possible phase equilibria involving phases based on aluminum, beryllium, and intermetallic compounds are proposed, and the types of strengthening of Al-Be alloys by an addition of a third element are classified.

  11. Acoustic tests of elastic and microplastic properties of VTiCr alloys

    NASA Astrophysics Data System (ADS)

    Chernov, V. M.; Rezvoushkin, A. V.; Kardashev, B. K.

    1996-10-01

    The non-linear acoustic properties of V10Ti5Cr alloy before and after proton irradiation (dose 2.2 × 10 14 p/cm 2) were investigated using a composite oscillator technique at longitudinal vibration frequencies of about 100 kHz. Acoustic parameters (decrement and resonance frequency) of the samples demonstrated noticeable amplitude dependencies of hysteretic type both in undeformed and deformed states. An unusual influence of plastical pre-staining on irradiated sample was found which resulted in small decreases in damping and increases in resonance frequency, and hence, of the elastic modulus. Damping in an irradiated sample was higher and its resonant frequency was lower as compared with a non-irradiated sample. This acoustic effect correlated with the results of microhardness and yield strength measurements. The experimental results are discussed in the framework of a model which predicts the creation by proton irradiation of defects which aid the motion of dislocations in V-alloys.

  12. Commercial food irradiation

    SciTech Connect

    Black, E.F.; Libby, L.M.

    1983-06-01

    Food irradiation is discussed. Irradiation exposes food to gamma rays from a cobalt-60 or a cesium-137 source, or to high-energy electrons emitted by an electron accelerator. A major advantage is that food can be packaged either before or after treatment. FDA regulations with regard to irradiation are discussed. Comments on an 'Advance Notice' on irradiation, published by the FDA in 1981 are summarized.

  13. Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications.

    SciTech Connect

    Hofman, G. L.

    1998-10-19

    Uranium alloys are candidates for the fuel phase in aluminum matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. Previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminum interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic {gamma}-phase during fabrication and irradiation, i.e., at temperatures at which {alpha}-U is the equilibrium phase. Transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research reactor operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analyzed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data.

  14. Inspection of irradiated P-7 fuel tubes

    SciTech Connect

    Peacock, H.B.; Sturcken, E.F.

    1980-08-20

    Mark 16 U-A1 alloy production fuel tubes and six special U{sub 3}O{sub 8}-A1 powder metallurgy (PM) test assemblies were successfully irradiated in P-7 reactor charge beginning December 1976. A year after irradiation, the outer surfaces were inspected under water in P-Area basin. Inspection showed that a black'' oxide had formed on the bottom {sup {approximately}}2/3 and flaked off in some areas for both the production and PM tubes. A small cladding defect was also observed on one PM outer tube near the bottom. Sections were cut from the tubes and metallographically examined in the SRL High Level Caves (HLC). This report gives results of the examinations. 8 refs., 9 figs., 1 tab.

  15. Irradiation creep of candidate materials for advanced nuclear plants

    NASA Astrophysics Data System (ADS)

    Chen, J.; Jung, P.; Hoffelner, W.

    2013-10-01

    In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2-8 × 10-6 dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

  16. Catalyst Alloys Processing

    NASA Astrophysics Data System (ADS)

    Tan, Xincai

    2014-10-01

    Catalysts are one of the key materials used for diamond formation at high pressures. Several such catalyst products have been developed and applied in China and around the world. The catalyst alloy most widely used in China is Ni70Mn25Co5 developed at Changsha Research Institute of Mining and Metallurgy. In this article, detailed techniques for manufacturing such a typical catalyst alloy will be reviewed. The characteristics of the alloy will be described. Detailed processing of the alloy will be presented, including remelting and casting, hot rolling, annealing, surface treatment, cold rolling, blanking, finishing, packaging, and waste treatment. An example use of the catalyst alloy will also be given. Industrial experience shows that for the catalyst alloy products, a vacuum induction remelt furnace can be used for remelting, a metal mold can be used for casting, hot and cold rolling can be used for forming, and acid pickling can be used for metal surface cleaning.

  17. Evolving Density and Static Mechanical Properties in Plutonium from Self-Irradiation

    SciTech Connect

    Chung, B W; Thompson, S R; Lema, K E; Hiromoto, D S; Ebbinghaus, B B

    2008-07-31

    Plutonium, because of its self-irradiation by alpha decay, ages by means of lattice damage and helium in-growth. These integrated aging effects result in microstructural and physical property changes. Because these effects would normally require decades to measure, studies are underway to assess the effects of extended aging on the physical properties of plutonium alloys by incorporating roughly 7.5 weight % of highly specific activity isotope {sup 238}Pu into the {sup 239}Pu metal to accelerate the aging process. This paper presents updated results of self-irradiation effects on {sup 238}Pu-enriched alloys measured by immersion density, dilatometry, and tensile tests. After nearly 90 equivalent years of aging, both the immersion density and dilatometry show that the enriched alloys continue to decreased in density by {approx}0.002% per year, without void swelling. Quasi-static tensile measurements show that the aging process increases the strength of plutonium alloys.

  18. Amorphous metal alloy

    DOEpatents

    Wang, R.; Merz, M.D.

    1980-04-09

    Amorphous metal alloys of the iron-chromium and nickel-chromium type have excellent corrosion resistance and high temperature stability and are suitable for use as a protective coating on less corrosion resistant substrates. The alloys are stabilized in the amorphous state by one or more elements of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The alloy is preferably prepared by sputter deposition.

  19. Low activation ferritic alloys

    DOEpatents

    Gelles, David S.; Ghoniem, Nasr M.; Powell, Roger W.

    1986-01-01

    Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

  20. Low activation ferritic alloys

    DOEpatents

    Gelles, D.S.; Ghoniem, N.M.; Powell, R.W.

    1985-02-07

    Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

  1. PLUTONIUM-ZIRCONIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  2. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part III: Phase stability during heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The phase stability in Fe-15Ni-13Cr alloys was investigated as a function of minor alloying additions after 4 MeV Ni ion irradiation at 948 K. The results showed that the stability of precipitate phases was dictated mainly by the defects produced by radiation damage and preferential segregation of Si and Ni at defects. In addition, radiation enhanced diffusion and cascade induced dissolution and mixing allowed kinetically sluggish phases to form rapidly under irradiation. These radiation effects caused an enhancement, retardation, or modification of thermal phases, and formation of new phases. The relative stability of precipitate phases varied sensitively with alloy composition. The roles of each alloying element on phase stability and the impact of radiation on the mechanisms of phase evolution were systematically studied and documented. The knowledge obtained from this work provides guidelines for designing alloys that lead to develop desired precipitate microstructures under irradiation.

  3. NICKEL-BASE ALLOY

    DOEpatents

    Inouye, H.; Manly, W.D.; Roche, T.K.

    1960-01-19

    A nickel-base alloy was developed which is particularly useful for the containment of molten fluoride salts in reactors. The alloy is resistant to both salt corrosion and oxidation and may be used at temperatures as high as 1800 deg F. Basically, the alloy consists of 15 to 22 wt.% molybdenum, a small amount of carbon, and 6 to 8 wt.% chromium, the balance being nickel. Up to 4 wt.% of tungsten, tantalum, vanadium, or niobium may be added to strengthen the alloy.

  4. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  5. METAPHIX-1 non destructive post irradiation examinations in the irradiated elements cell at Phenix

    SciTech Connect

    Breton, Laurent; Masson, M.; Garces, E.; Desjardins, S.; Fontaine, B.; Lacroix, B.; Martella, T.; Loubet, L.; Ohta, H.; Yokoo, T.; Ougier, M.; Glatz, J.P.

    2007-07-01

    Central Research Institute of Electric Power Industry (CRIEPI) has been developing minor actinide (MA) transmutation technology in homogeneous loading mode by use of metal fuel fast reactors in cooperation with Institute for Transuranium Elements (ITU) and Commissariat a l'Energie Atomique (CEA). Fast reactor metal fuel pins of Uranium- Plutonium-Zirconium (U-Pu-Zr) alloy containing 2 wt% MAs and 2 wt% rare earth elements (REs), 5 wt% MAs, and 5 wt% MAs and 5 wt% REs were irradiated in the PHENIX French fast reactor as METAPHIX experiments. In these METAPHIX experiments, three rigs each consisting of three metal fuel experimental pins and sixteen oxide fuel driver pins were irradiated. The target burnup of the three rigs is 2.4 at%, 7 at% and 11 at% which corresponds to 120, 360 and 600 equivalent full power days (EFPD) in terms of irradiation periods, respectively. The low burnup rig of 2.4 at%, METAPHIX-1, was discharged from the core in August 2004. After cooling, the non-destructive post irradiation examinations (PIEs) of the rig (visual examination, measurement of rig length and deformation) and of the metal fuel pins (visual examination, measurement of pin length and deformation, {gamma}-spectrometry and neutron radiography) were conducted in the Irradiated Elements Cell (IEC) at PHENIX. (authors)

  6. Precipitate stability in neutron-irradiated Zircaloy-4

    NASA Astrophysics Data System (ADS)

    Yang, W. J. S.

    1988-09-01

    Zircaloy-4, a zirconium-base alloy used extensively as cladding and core structural materials in water-cooled nuclear reactors, was examined by transmission electron microscopy, after neutron irradiation and postirradiation annealing. Phase instabilities found during irradiation at 561 K include the amorphous transformation and the dissolution of the intermetallic Zr(Fe,Cr) 2. The α-matrix is driven toward a single phase solid solution as the neutron fluence increases. This is evidenced by the continuous dissolution of the precipitate without precipitation of any new phase during irradiation. During postirradiation annealing at 833 K, solute Fe precipitates out particularly at the grain boundaries as Zr-Fe zeta-phase. Recrystallization of the amorphous precipitates occurs at a postirradiation annealing temperature of 1023 K. In general, the observed phenomena of amorphous transformation, precipitate dissolution, reprecipitation and recrystallization reflect the complex solute-point defect interactions in the α-matrix. The continuous solute dissolution during irradiation is expected to have a potential effect on irradiation growth, creep and corrosion properties of the alloy.

  7. Revised ANL-reported tensile data for V-Ti and V-Cr-Ti alloys

    SciTech Connect

    Billone, M.C.

    1997-08-01

    The tensile for all irradiated vanadium alloy samples and several unirradiated vanadium alloys tested at Argonne National Laboratory (ANL) have been critically reviewed and revised, as necessary. The review and revision are based on re-analyzing the original load-displacement strip-chart recording using a methodology consistent with current ASTM standards. No significant difference has been found between the newly-revised and previously-reported values of yield strength (YS) and ultimate tensile strength (UTS). However, by correctly subtracting the non-gauge-length displacement and linear gauge-length displacement from the total cross-head displacement, the uniform elongation (UE) of the gauge length decreases by 4-9% strain and the total elongation (TE) of the gauge length decreases by 1-7% strain. These differences are more significant for lower-ductility irradiated alloys than for higher-ductility alloys.

  8. Characterization of Two ODS Alloys: Chromium-18 ODS and Chromium-9 ODS

    NASA Astrophysics Data System (ADS)

    Goddard, Julianne

    ODS alloys, or oxide dispersion strengthened alloys, are made from elemental or pre-alloyed metal powders mechanically alloyed with oxide powders in a high-energy attributor mill, and then consolidated by either hot isostatic pressing or hot extrusion causing the production of nanometer scale oxide and carbide particles within the alloy matrix; crystalline properties such as creep strength, ductility, corrosion resistance, tensile strength, swelling resistance, and resistance to embrittlement are all observed to be improved by the presence of nanoparticles in the matrix. The presented research uses various methods to observe and characterize the microstructural and microchemical properties of two experimental ODS alloys, 18Cr ODS and 9Cr ODS. The results found aid in assessing the influence of chemical and structural variations on the effectiveness of the alloy, and further aid in the optimization of these advanced alloys for future use in nuclear cladding and structural applications in Generation IV nuclear reactors. Characterization of these alloys has been conducted in order to identify the second-phase small precipitates through FESEM, TEM, EDS, Synchrotron X-ray diffraction analysis, and CuKalpha XRD analysis of bulk samples and of nanoparticles after extraction from the alloy matrix. Comparison of results from these methods allows further substantiation of the accuracy of observed nanoparticle composition and identification. Also, TEM samples of the two alloys have been irradiated in-situ with 1 MeV Kr and 300 keV Fe ions to various doses and temperatures at the IVEM-Tandem TEM at Argonne National Laboratory and post-irradiated characterization has been conducted and compared to the pre-irradiated characterization results in order to observe the microstructural and microchemical evolution of nanoparticles under irradiation. Overall in the as-received state, the initial Y2O3 is not found anymore and in addition to oxide particles the alloys contain carbides

  9. Subtask 12F2: Microstructural evolution of V-4Cr-4Ti during neutron irradiation

    SciTech Connect

    Chung, H.M.; Gazda, J.; Loomis, B.A.

    1995-03-01

    The objective of this work is to characterize the microstructural evolution of V-4Cr-4Ti alloy during irradiation by fast neutrons, and thereby to provide a better understanding of long-term performance of the alloy under fusion conditions. Microstructural evolution of V-4Cr-4Ti, an alloy recently shown to exhibit excellent tensile and creep properties, virtual immunity to irradiation embrittlement, and good resistance to swelling, was characterized after irradiation in a lithium environment in the Fast Flux Test Facility (FFTF) (a sodium-cooled fast reactor located in Richland, Washington) at 420, 520, and 600{degrees}C to 24-34 dpa. The primary feature of microstructural evolution during irradiation at 520 and 600{degrees}C was high-density formation of ultrafine Ti{sub 5}Si{sub 3} precipitates and short dislocations. For irradiation at 420{degrees}C, precipitation of Ti{sub 5}Si{sub 3} was negligible, and {open_quotes}black-dot{close_quotes} defects and dislocations were observed in significantly higher densities. In spite of their extremely high densities, neither the {open_quotes}black-dot{close_quotes} defects nor Ti{sub 5}Si{sub 3} precipitates are overly detrimental to ductility and toughness of the alloy, yet they very effectively suppress irradiation-induced swelling. Therefore, these features, normally observed in V-base alloys containing Ti and Si, are considered stable. Unstable microstructural modifications that are likely to degrade mechanical properties significantly were not observed, e.g., irradiation-induced formation of fine oxides, carbides, nitrides, or Cr-rich clusters. 18 refs., 4 figs., 1 tab.

  10. The role of dislocation channeling in IASCC initiation of neutron irradiated austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale Jennings

    The objective of this study was to understand the role of dislocation channeling in the initiation of irradiation-assisted stress corrosion cracking (IASCC) of neutron irradiated austenitic stainless steel using a novel four-point bend test. Stainless steels used in this study were irradiated in the BOR-60 fast reactor at 320 °C, and included a commercial purity 304L stainless steel irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity stainless steels, Fe-18Cr-12Ni and Fe-18Cr-25Ni, irradiated to ~10 dpa. The four-point bend test produced the same relative IASCC susceptibility as constant extension rate tensile (CERT) experiments performed on the same irradiated alloys in boiling water reactor normal water chemistry. The cracking susceptibility of the CP 304L alloy was high at all irradiation dose levels, enhanced by the presence of MnS inclusions in the alloy microstructure, which dissolve in the NWC environment. Dissolution of the MnS inclusion results in formation of an oxide cap that occludes the inclusion site, creating a crevice condition with a high propensity for crack initiation. Crack initiation at these locations was induced by stress concentration at the intersecting grain boundary, resulting from the intersection of a discontinuous dislocation channels (DC). Stress to initiate an IASCC crack decreased with dose due earlier DC initiation. The HP Fe-18Cr-12Ni alloy had low susceptibility to IASCC, while the high Ni alloy exhibited no cracking susceptibility. The difference in susceptibility among these conditions was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC -- grain boundary intersections.

  11. The effects of oxide evolution on mechanical properties in proton- and neutron-irradiated Fe-9%Cr ODS steel

    NASA Astrophysics Data System (ADS)

    Swenson, M. J.; Dolph, C. K.; Wharry, J. P.

    2016-10-01

    The objective of this study is to evaluate the effect of irradiation on the strengthening mechanisms of a model Fe-9%Cr oxide dispersion strengthened steel. The alloy was irradiated with protons or neutrons to a dose of 3 displacements per atoms at 500 °C. Nanoindentation was used to measure strengthening due to irradiation, with neutron irradiation causing a greater increase in yield strength than proton irradiation. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography (APT). Cluster analysis reveals solute migration from the Y-Ti-O-rich nanoclusters to the surrounding matrix after both irradiations, though the effect is more pronounced in the neutron-irradiated specimen. Because the dissolved oxygen atoms occupy interstitial sites in the iron matrix, they contribute significantly to solid solution strengthening. The dispersed barrier hardening model relates microstructure evolution to the change in yield strength, but is only accurate if solid solution contributions to strengthening are considered simultaneously.

  12. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel.

    SciTech Connect

    Chung, H. M.; Ruther, W. E.; Strain, R. V.; Shack, W. J.; Karlsen, T. M.

    1999-10-26

    Slow-strain-rate tensile (SSRT) tests were conducted on model austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup 2} and {approx} 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. At {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, a high-purity heat of Type 316L SS that contains a very low concentration of Si exhibited the highest susceptibility to IGSCC. In unirradiated state, Types 304 and 304L SS did not exhibit a systematic effect of Si content on alloy strength. However, at {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, yield and maximum strengths decreased significantly as Si content was increased to >0.9 wt.%. Among alloys that contain low concentrations of C and N, ductility and resistance to TGSCC and IGSCC were significantly greater for alloys with >0.9 wt.% Si than for alloys with <0.47 wt.% Si. Initial data at {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} were also consistent with the beneficial effect of high Si content. This indicates that to delay onset of and reduce susceptibility to irradiation-assisted stress corrosion cracking (IASCC), at least at low fluence levels, it is helpful to ensure a certain minimum concentration of Si. High concentrations of Cr were also beneficial; alloys that contain <15.5 wt.% Cr exhibited greater susceptibility to IASCC than alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  13. Aluminium. II - A review of deformation properties of high purity aluminium and dilute aluminium alloys.

    NASA Technical Reports Server (NTRS)

    Reed, R. P.

    1972-01-01

    The elastic and plastic deformation behavior of high-purity aluminum and of dilute aluminum alloys is reviewed. Reliable property data, including elastic moduli, elastic coefficients, tensile, creep, fatigue, hardness, and impact are presented. Single crystal tensile results are discussed. Rather comprehensive reference lists, containing publications of the past 20 years, are included for each of the above categories. Defect structures and mechanisms responsible for mechanical behavior are presented. Strengthening techniques (alloys, cold work, irradiation, quenching, composites) and recovery are briefly reviewed.

  14. Copper-tantalum alloy

    DOEpatents

    Schmidt, Frederick A.; Verhoeven, John D.; Gibson, Edwin D.

    1986-07-15

    A tantalum-copper alloy can be made by preparing a consumable electrode consisting of an elongated copper billet containing at least two spaced apart tantalum rods extending longitudinally the length of the billet. The electrode is placed in a dc arc furnace and melted under conditions which co-melt the copper and tantalum to form the alloy.

  15. Cesium iodide alloys

    DOEpatents

    Kim, H.E.; Moorhead, A.J.

    1992-12-15

    A transparent, strong CsI alloy is described having additions of monovalent iodides. Although the preferred iodide is AgI, RbI and CuI additions also contribute to an improved polycrystalline CsI alloy with outstanding multispectral infrared transmittance properties. 6 figs.

  16. Ductile transplutonium metal alloys

    DOEpatents

    Conner, William V.

    1983-01-01

    Alloys of Ce with transplutonium metals such as Am, Cm, Bk and Cf have properties making them highly suitable as sources of the transplutonium element, e.g., for use in radiation detector technology or as radiation sources. The alloys are ductile, homogeneous, easy to prepare and have a fairly high density.

  17. Nickel base coating alloy

    NASA Technical Reports Server (NTRS)

    Barrett, C. A. (Inventor); Lowell, C. E. (Inventor)

    1986-01-01

    Zirconium is added to a Ni-30 Al (beta) intermetallic alloy in the range of 0.05 w/o to 0.25 w/o. This addition is made during melting or by using metal powders. The addition of zirconium improves the cyclic oxidation resistance of the alloys at temperatures above 1100 C.

  18. PLUTONIUM-CERIUM ALLOY

    DOEpatents

    Coffinberry, A.S.

    1959-01-01

    An alloy is presented for use as a reactor fuel. The binary alloy consists essentially of from about 5 to 90 atomic per cent cerium and the balance being plutonium. A complete phase diagram for the cerium--plutonium system is given.

  19. Neutron Absorbing Alloys

    DOEpatents

    Mizia, Ronald E.; Shaber, Eric L.; DuPont, John N.; Robino, Charles V.; Williams, David B.

    2004-05-04

    The present invention is drawn to new classes of advanced neutron absorbing structural materials for use in spent nuclear fuel applications requiring structural strength, weldability, and long term corrosion resistance. Particularly, an austenitic stainless steel alloy containing gadolinium and less than 5% of a ferrite content is disclosed. Additionally, a nickel-based alloy containing gadolinium and greater than 50% nickel is also disclosed.

  20. Aluminum battery alloys

    DOEpatents

    Thompson, D.S.; Scott, D.H.

    1984-09-28

    Aluminum alloys suitable for use as anode structures in electrochemical cells are disclosed. These alloys include iron levels higher than previously felt possible, due to the presence of controlled amounts of manganese, with possible additions of magnesium and controlled amounts of gallium.

  1. Aluminum battery alloys

    DOEpatents

    Thompson, David S.; Scott, Darwin H.

    1985-01-01

    Aluminum alloys suitable for use as anode structures in electrochemical cs are disclosed. These alloys include iron levels higher than previously felt possible, due to the presence of controlled amounts of manganese, with possible additions of magnesium and controlled amounts of gallium.

  2. Ductile transplutonium metal alloys

    DOEpatents

    Conner, W.V.

    1981-10-09

    Alloys of Ce with transplutonium metals such as Am, Cm, Bk and Cf have properties making them highly suitable as souces of the transplutonium element, e.g., for use in radiation detector technology or as radiation sources. The alloys are ductile, homogeneous, easy to prepare and have a fairly high density.

  3. Method for improving performance of irradiated structural materials

    DOEpatents

    Megusar, Janez; Harling, Otto K.; Grant, Nicholas J.

    1989-01-01

    Method for extending service life of nuclear reactor components prepared from ductile, high strength crystalline alloys obtained by devitrification of metallic glasses. Two variations of the method are described: (1) cycling the temperature of the nuclear reactor between the operating temperature which leads to irradiation damage and a l The U.S. Government has rights in this invention by virtue of Department of Energy, Office of Fusion Energy, Grant No. DE-AC02-78ER-10107.

  4. Ultrahigh temperature intermetallic alloys

    SciTech Connect

    Brady, M.P.; Zhu, J.H.; Liu, C.T.; Tortorelli, P.F.; Wright, J.L.; Carmichael, C.A.; Walker, L.R.

    1997-12-01

    A new family of Cr-Cr{sub 2}X based alloys with fabricability, mechanical properties, and oxidation resistance superior to previously developed Cr-Cr{sub 2}Nb and Cr-Cr{sub 2}Zr based alloys has been identified. The new alloys can be arc-melted/cast without cracking, and exhibit excellent room temperature and high-temperature tensile strengths. Preliminary evaluation of oxidation behavior at 1100 C in air indicates that the new Cr-Cr{sub 2}X based alloys form an adherent chromia-based scale. Under similar conditions, Cr-Cr{sub 2}Nb and Cr-Cr{sub 2}Zr based alloys suffer from extensive scale spallation.

  5. Immersion studies on candidate container alloys for the Tuff Repository

    SciTech Connect

    Beavers, J.A.; Durr, C.L.

    1991-05-01

    Cortest Columbus Technologies (CC Technologies) is investigating the long-term performance of container materials used for high-level radioactive waste packages. This information is being developed for the Nuclear Regulatory Commission to aid in their assessment of the Department of Energy`s application to construct a geologic repository for disposal of high-level radioactive waste. This report summarizes the results of exposure studies performed on two copper-base and two Fe-Cr-Ni alloys in simulated Tuff Repository conditions. Testing was performed at 90{degrees}C in three environments; simulated J-13 well water, and two environments that simulated the chemical effects resulting from boiling and irradiation of the groundwater. Creviced specimens and U-bends were exposed to liquid, to vapor above the condensed phase, and to alternate immersion. A rod specimen was used to monitor corrosion at the vapor-liquid interface. The specimens were evaluated by electrochemical, gravimetric, and metallographic techniques following approximately 2000 hours of exposure. Results of the exposure tests indicated that all four alloys exhibited acceptable general corrosion rates in simulated J-13 well water. These rates decreased with time. Incipient pitting was observed under deposits on Alloy 825 and pitting was observed on both Alloy CDA 102 and Alloy CDA 715 in the simulated J-13 well water. No SCC was observed in U-bend specimens of any of the alloys in simulated J-13 well water. 33 refs., 48 figs., 23 tabs.

  6. Grain boundary migration induced segregation in V-Cr-Ti alloys

    SciTech Connect

    Gelles, D.S.; Ohnuki, S.; Takahashi, H.

    1996-10-01

    Analytical electron microscopy results are reported for a series of vanadium alloys irradiated in the HFIR JP23 experiment at 500{degrees}C. Alloys were V-5Cr-5Ti and pure vanadium which are expected to have transmuted to V-15Cr-5Ti and V-10Cr following irradiation. Analytical microscopy confirmed the expected transmutation occurred and showed redistribution of Cr and Ti resulting from grain boundary migration in V-5Cr-5Ti, but in pure V, segregation was reduced and no clear trends as a function of position near a boundary were identified.

  7. Characterization of nuclear transmutations in materials irradiated test facilities

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1994-05-01

    This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available, fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V-4Cr-4Ti), a reduced activation martensitic steel (Fe-9Cr-2WVTa), a high conductivity copper alloy (Cu-Cr-Zr), and the SiC compound. The fission irradiation facilities considered include the EBR-II fast reactor, and two high flux mixed spectrum reactors, HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The transmutation and dpa rates that occur in these test reactors are compared with the calculated transmutation and dpa rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions [(n,x,{alpha}) and (n,xp)] are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from the ones predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time.

  8. The effect of oxygen on void stability in ion-irradiated steel

    NASA Astrophysics Data System (ADS)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  9. Validation of the shear punch-tensile correlation technique using irradiated materials

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Toloczko, M.B.; Hamilton, M.L.

    1998-03-01

    It was recently demonstrated that tensile data could be successfully related to shear punch data obtained on transmission electron microscopy (TEM) discs for a variety of irradiated alloys exhibiting yield strengths that ranged from 100 to 800 MPa. This implies that the shear punch test might be a viable alternative for obtaining tensile properties using a TEM disk, which is much smaller than even the smallest miniature tensile specimens, especially when irradiated specimens are not available or when they are too radioactive to handle easily. The majority of the earlier tensile-shear punch correlation work was done using a wide variety of unirradiated materials. The current work extends this correlation effort to irradiated materials and demonstrates that the same relationships that related shear punch tests remain valid for irradiated materials. Shear punch tests were performed on two sets of specimens. In the first group, three simple alloys from the {sup 59}Ni isotopic doping series in the solution annealed and cold worked conditions were irradiated at temperatures ranging from 365 to 495 C in the Fast Flux Test Facility. The corresponding tensile data already existed for tensile specimens fabricated from the same raw materials and irradiated side-by-side with the disks. In the second group, three variants of 316 stainless steel were irradiated in FFTF at 5 temperatures between 400 and 730 C to doses ranging from 12.5 to 88 dpa. The specimens were in the form of both TEM and miniature tensile specimens and were irradiated side-by-side.

  10. THORIUM-SILICON-BERYLLIUM ALLOYS

    DOEpatents

    Foote, F.G.

    1959-02-10

    Th, Si, anol Bt alloys where Be and Si are each present in anmounts between 0.1 and 3.5% by weight and the total weight per cent of the minor alloying elements is between 1.5 and 4.5% are discussed. These ternary alloys show increased hardness and greater resistant to aqueous corrosion than is found in pure Th, Th-Si alloys, or Th-Be alloys.

  11. Present Status of Vanadium Alloys for Fusion Applications

    SciTech Connect

    Muroga, Takeo; Chen, J. M.; Chernov, V. M.; Kurtz, Richard J.; Le Flem, M.

    2014-12-01

    Vanadium alloys are advanced options for low activation structural materials. After more than two decades of research, V-4Cr-4Ti has been emerged as the leading candidate, and technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. Notable progress has been made in fabricating alloy products and weld joints without degradation of properties. Various efforts are also being made to improve high temperature strength and creep-rupture resistance, low temperature ductility after irradiation, and corrosion resistance in blanket conditions. Future research should focus on clarifying remaining uncertainty in the operating temperature window of V-4Cr-4Ti for application to near to middle term fusion blanket systems, and on further exploration of advanced materials for improved performance for longer-term fusion reactor systems.

  12. Magnesium silicide intermetallic alloys

    NASA Astrophysics Data System (ADS)

    Li, Gh.; Gill, H. S.; Varin, R. A.

    1993-11-01

    Methods of induction melting an ultra-low-density magnesium silicide (Mg2Si) intermetallic and its alloys and the resulting microstructure and microhardness were studied. The highest quality ingots of Mg2Si alloys were obtained by triple melting in a graphite crucible coated with boron nitride to eliminate reactivity, under overpressure of high-purity argon (1.3 X 105 Pa), at a temperature close to but not exceeding 1105 °C ± 5 °C to avoid excessive evaporation of Mg. After establishing the proper induction-melting conditions, the Mg-Si binary alloys and several Mg2Si alloys macroalloyed with 1 at. pct of Al, Ni, Co, Cu, Ag, Zn, Mn, Cr, and Fe were induction melted and, after solidification, investigated by optical microscopy and quantitative X-ray energy dispersive spectroscopy (EDS). Both the Mg-rich and Si-rich eutectic in the binary alloys exhibited a small but systematic increase in the Si content as the overall composition of the binary alloy moved closer toward the Mg2Si line compound. The Vickers microhardness (VHN) of the as-solidified Mg-rich and Si-rich eutectics in the Mg-Si binary alloys decreased with increasing Mg (decreasing Si) content in the eutectic. This behavior persisted even after annealing for 75 hours at 0.89 pct of the respective eutectic temperature. The Mg-rich eutectic in the Mg2Si + Al, Ni, Co, Cu, Ag, and Zn alloys contained sections exhibiting a different optical contrast and chemical composition than the rest of the eutectic. Some particles dispersed in the Mg2Si matrix were found in the Mg2Si + Cr, Mn, and Fe alloys. The EDS results are presented and discussed and compared with the VHN data.

  13. Influence of chemical disorder on energy dissipation and defect evolution in advanced alloys

    DOE PAGES

    Zhang, Yanwen; Jin, Ke; Xue, Haizhou; Lu, Chenyang; Olsen, Raina J.; Beland, Laurent K.; Ullah, Mohammad W.; Zhao, Shijun; Bei, Hongbin; Aidhy, Dilpuneet S.; et al

    2016-08-01

    We report that historically, alloy development with better radiation performance has been focused on traditional alloys with one or two principal element(s) and minor alloying elements, where enhanced radiation resistance depends on microstructural or nanoscale features to mitigate displacement damage. In sharp contrast to traditional alloys, recent advances of single-phase concentrated solid solution alloys (SP-CSAs) have opened up new frontiers in materials research. In these alloys, a random arrangement of multiple elemental species on a crystalline lattice results in disordered local chemical environments and unique site-to-site lattice distortions. Based on closely integrated computational and experimental studies using a novel setmore » of SP-CSAs in a face-centered cubic structure, we have explicitly demonstrated that increasing chemical disorder can lead to a substantial reduction in electron mean free paths, as well as electrical and thermal conductivity, which results in slower heat dissipation in SP-CSAs. The chemical disorder also has a significant impact on defect evolution under ion irradiation. Considerable improvement in radiation resistance is observed with increasing chemical disorder at electronic and atomic levels. Finally, the insights into defect dynamics may provide a basis for understanding elemental effects on evolution of radiation damage in irradiated materials and may inspire new design principles of radiation-tolerant structural alloys for advanced energy systems.« less

  14. Comminuting irradiated ferritic steel

    DOEpatents

    Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

    1985-01-01

    Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

  15. MASSIVE LEAKAGE IRRADIATOR

    DOEpatents

    Wigner, E.P.; Szilard, L.; Christy, R.F.; Friedman, F.L.

    1961-05-30

    An irradiator designed to utilize the neutrons that leak out of a reactor around its periphery is described. It avoids wasting neutron energy and reduces interference with the core flux to a minimum. This is done by surrounding all or most of the core with removable segments of the material to be irradiated within a matrix of reflecting material.

  16. Perspective on food irradiation

    SciTech Connect

    Not Available

    1987-02-01

    Recent US Food and Drug Administration approval of irradiation treatment for fruit, vegetables and pork has stimulated considerable discussion in the popular press on the safety and efficacy of irradiation processing of food. This perspective is designed to summarize the current scientific information available on this issue.

  17. TUNGSTEN BASE ALLOYS

    DOEpatents

    Schell, D.H.; Sheinberg, H.

    1959-12-15

    A high-density quaternary tungsten-base alloy having high mechanical strength and good machinability composed of about 2 wt.% Ni, 3 wt.% Cu, 5 wt.% Pb, and 90wt.% W is described. This alloy can be formed by the powder metallurgy technique of hot pressing in a graphite die without causing a reaction between charge and the die and without formation of a carbide case on the final compact, thereby enabling re-use of the graphite die. The alloy is formable at hot- pressing temperatures of from about 1200 to about 1350 deg C. In addition, there is little component shrinkage, thereby eliminating the necessity of subsequent extensive surface machining.

  18. Irradiation Creep in Graphite

    SciTech Connect

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  19. Effect of irradiation temperature and strain rate on the mechanical properties of V-4Cr-4Ti irradiated to low doses in fission reactors

    SciTech Connect

    Zinkle, S.J.; Snead, L.L.; Rowcliffe, A.F.; Alexander, D.J.; Gibson, L.T.

    1998-09-01

    Tensile tests performed on irradiated V-(3-6%)Cr-(3-6%)Ti alloys indicate that pronounced hardening and loss of strain hardening capacity occurs for doses of 0.1--20 dpa at irradiation temperatures below {approximately}330 C. The amount of radiation hardening decreases rapidly for irradiation temperatures above 400 C, with a concomitant increase in strain hardening capacity. Low-dose (0.1--0.5 dpa) irradiation shifts the dynamic strain aging regime to higher temperatures and lower strain rates compared to unirradiated specimens. Very low fracture toughness values were observed in miniature disk compact specimens irradiated at 200--320 C to {approximately}1.5--15 dpa and tested at 200 C.

  20. Irradiation response and stability of nanoporous materials

    SciTech Connect

    Fu, Engang; Wang, Yongqiang; Serrano De Caro, Magdalena; Caro, Jose A.; Zepeda-Ruiz, L; Bringa, E.; Nastasi, Mike; Baldwin, Jon K.

    2012-08-28

    Nanoporous materials consist of a regular organic or inorganic framework supporting a regular, porous structure. Pores are by definition roughly in the nanometre range, that is between 0.2 nm and 100 nm. Nanoporous materials can be subdivided into 3 categories (IUPAC): (1) Microporous materials - 0.2-2 nm; (2) Mesoporous materials - 2-50 nm; and (3) Macroporous materials - 50-1000 nm. np-Au foams were successfully synthesized by de-alloying process. np-Au foams remain porous structure after Ne ion irradiation to 1 dpa. Stacking Fault Tetrahedra (SFTs) were observed in RT irradiated np-Au foams under the highest and intermediate fluxes, but not under the lowest flux. SFTs were not observed in LNT irradiated np-Au foams under all fluxes. The vacancy diffusivity in Au at RT is high enough so that the vacancies have enough time to agglomerate and then collapse to form SFTs. The high ion flux creates more damage per unit time; vacancies don't have enough time to diffuse or recombine. As a result, SFTs were formed at high ion fluxes.

  1. Irradiation Environment of the Materials Test Station

    SciTech Connect

    Pitcher, Eric John

    2012-06-21

    Conceptual design of the proposed Materials Test Station (MTS) at the Los Alamos Neutron Science Center (LANSCE) is now complete. The principal mission is the irradiation testing of advanced fuels and materials for fast-spectrum nuclear reactor applications. The neutron spectrum in the fuel irradiation region of MTS is sufficiently close to that of fast reactor that MTS can match the fast reactor fuel centerline temperature and temperature profile across a fuel pellet. This is an important characteristic since temperature and temperature gradients drive many phenomena related to fuel performance, such as phase stability, stoichiometry, and fission product transport. The MTS irradiation environment is also suitable in many respects for fusion materials testing. In particular, the rate of helium production relative to atomic displacements at the peak flux position in MTS matches well that of fusion reactor first wall. Nuclear transmutation of the elemental composition of the fusion alloy EUROFER97 in MTS is similar to that expected in the first wall of a fusion reactor.

  2. Advanced TEM characterization of oxide nanoparticles in ODS Fe–12Cr–5Al alloys

    DOE PAGES

    Unocic, Kinga A.; Pint, Bruce A.; Hoelzer, David T.

    2016-07-11

    For oxide nanoparticles present in three oxide-dispersion-strengthened (ODS) Fe–12Cr–5Al alloys containing additions of (1) Y2O3 (125Y), (2) Y2O3 + ZrO2 (125YZ), and (3) Y2O3 + HfO2 (125YH), were investigated using transmission and scanning transmission electron microscopy. Furthermore, in all three alloys nano-sized (<3.5 nm) oxide particles distributed uniformly throughout the microstructure were characterized using advanced electron microscopy techniques. In the 125Y alloy, mainly Al2O3 and yttrium–aluminum garnet (YAG) phases (Y3Al5O12) were present, while in the 125YZ alloy, additional Zr(C,N) precipitates were identified. The 125YH alloy had the most complex precipitation sequence whereby in addition to the YAG and Al2O3 phases,more » Hf(C,N), Y2Hf2O7, and HfO2 precipitates were also found. The presence of HfO2 was mainly due to the incomplete incorporation of HfO2 powder during mechanical alloying of the 125YH alloy. The alloy having the highest total number density of the oxides, the smallest grain size, and the highest Vickers hardness was the 125YZ alloy indicating, that Y2O3 + ZrO2 additions had the strongest effect on grain size and tensile properties. Finally, high-temperature mechanical testing will be addressed in the near future, while irradiation studies are underway to investigate the irradiation resistance of these new ODS FeCrAl alloys.« less

  3. Radiation-Damage in Molybdenum-Rhenium Alloys for Space Reactor Applications

    SciTech Connect

    Busby, Jeremy T; Leonard, Keith J; Zinkle, Steven J

    2007-01-01

    Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41% to 47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the mechanical properties of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ~ 0.7 dpa at 1073, 1223, and 1373 K and ~1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 x 10-3 s-1 in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 hours at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined.

  4. The nano-particle dispersion strengthening of V-4Cr-4Ti alloys for high temperature application in fusion reactors

    NASA Astrophysics Data System (ADS)

    Zheng, Pengfei; Chen, Jiming; Xu, Zengyu; Duan, Xuru

    2013-10-01

    V-4Cr-4Ti was identified as an attractive structural material for Li blanket in fusion reactors. However, both high temperature and irradiation induced degradation are great challenges for this material. It was thought that the nano-particles with high thermal stability can efficiently strengthen the alloy at elevated temperatures, and accommodate the irradiation induced defects at the boundaries. This study is a starting work aiming at improving the creep resistance and reducing the irradiation induced degradation for V-4Cr-4Ti alloy. Currently, we focus on the preparation of some comparative nano-particle dispersion strengthened V-4Cr-4Ti alloys. A mechanical alloying (MA) route is used to fabricate yttrium and carbides added V-4Cr-4Ti alloys. Nano-scale yttria, carbides and other possible particles have a combined dispersion-strengthening effect on the matrices of these MA-fabricated V-4Cr-4Ti alloys. High-temperature annealing is carried out to stabilize the optimized nano-particles. Mechanical properties are tested. Microstructures of the MA-fabricated V-4Cr-4Ti alloys with yttrium and carbide additions are characterized. Based on these results, the thermal stability of different nano-particle agents are classified. ITER related China domestic project 2011GB108007.

  5. Electroplating on titanium alloy

    NASA Technical Reports Server (NTRS)

    Lowery, J. R.

    1971-01-01

    Activation process forms adherent electrodeposits of copper, nickel, and chromium on titanium alloy. Good adhesion of electroplated deposits is obtained by using acetic-hydrofluoric acid anodic activation process.

  6. Alloy Selection System

    SciTech Connect

    2001-02-01

    Software will Predict Corrosion Rates to Improve Productivity in the Chemical Industry. Many aspects of equipment design and operation are influenced by the choice of the alloys used to fabricate process equipment.

  7. Metallographic analysis of irradiated RERTR-3 fuel test specimens.

    SciTech Connect

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-11-08

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date.

  8. PLUTONIUM-URANIUM ALLOY

    DOEpatents

    Coffinberry, A.S.; Schonfeld, F.W.

    1959-09-01

    Pu-U-Fe and Pu-U-Co alloys suitable for use as fuel elements tn fast breeder reactors are described. The advantages of these alloys are ease of fabrication without microcracks, good corrosion restatance, and good resistance to radiation damage. These advantages are secured by limitation of the zeta phase of plutonium in favor of a tetragonal crystal structure of the U/sub 6/Mn type.

  9. Disk Alloy Development

    NASA Technical Reports Server (NTRS)

    Gabb, Tim; Gayda, John; Telesman, Jack

    2001-01-01

    The advanced powder metallurgy disk alloy ME3 was designed using statistical screening and optimization of composition and processing variables in the NASA HSR/EPM disk program to have extended durability at 1150 to 1250 "Fin large disks. Scaled-up disks of this alloy were produced at the conclusion of this program to demonstrate these properties in realistic disk shapes. The objective of the UEET disk program was to assess the mechanical properties of these ME3 disks as functions of temperature, in order to estimate the maximum temperature capabilities of this advanced alloy. Scaled-up disks processed in the HSR/EPM Compressor / Turbine Disk program were sectioned, machined into specimens, and tested in tensile, creep, fatigue, and fatigue crack growth tests by NASA Glenn Research Center, in cooperation with General Electric Engine Company and Pratt & Whitney Aircraft Engines. Additional sub-scale disks and blanks were processed and tested to explore the effects of several processing variations on mechanical properties. Scaled-up disks of an advanced regional disk alloy, Alloy 10, were used to evaluate dual microstructure heat treatments. This allowed demonstration of an improved balance of properties in disks with higher strength and fatigue resistance in the bores and higher creep and dwell fatigue crack growth resistance in the rims. Results indicate the baseline ME3 alloy and process has 1300 to 1350 O F temperature capabilities, dependent on detailed disk and engine design property requirements. Chemistry and process enhancements show promise for further increasing temperature capabilities.

  10. Radiation hardening and deformation behavior of irradiated ferritic-martensitic steels

    SciTech Connect

    Robertson, J.P.; Klueh, R.L.; Rowcliffe, A.F.; Shiba, K.

    1998-03-01

    Tensile data from several 8--12% Cr alloys irradiated in the High Flux Isotope Reactor (HFIR) to doses up to 34 dpa at temperatures ranging from 90 to 600 C are discussed in this paper. One of the critical questions surrounding the use of ferritic-martensitic steels in a fusion environment concerns the loss of uniform elongation after irradiation at low temperatures. Irradiation and testing at temperatures below 200--300 C results in uniform elongations less than 1% and stress-strain curves in which plastic instability immediately follows yielding, implying dislocation channeling and flow localization. Reductions in area and total elongations, however, remain high.

  11. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

    SciTech Connect

    Chen, Y.; Yang, Yong; Huang, Yina; Allen, T.; Alexandreanu, B.; Natesan, K.

    2012-11-01

    As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10-6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

  12. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment

    SciTech Connect

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1995-04-01

    The objective of this work is to determine the effects of neutron irradiation on the mechanical properties of austenitic stainless steel alloys. In this experiment, the spectrum has been tailored to reduce the thermal neutron flux and achieve a He/dpa level near that expected in a fusion reactor.

  13. Test reactor irradiation coordination

    SciTech Connect

    Heartherly, D.W.; Siman Tov, I.I.; Sparks, D.W.

    1995-10-01

    This task was established to supply and coordinate irradiation services needed by NRC contractors other than ORNL. These services include the design and assembly of irradiation capsules as well as arranging for their exposure, disassembly, and return of specimens. During this period, the final design of the facility and specimen baskets was determined through an iterative process involving the designers and thermal analysts. The resulting design should permit the irradiation of all test specimens to within 5{degrees}C of their desired temperature. Detailing of all parts is ongoing and should be completed during the next reporting period. Procurement of the facility will also be initiated during the next review period.

  14. Alaskan Commodities Irradiation Project

    SciTech Connect

    Zarling, J.P.; Swanson, R.B.; Logan, R.R.; Das, D.K.; Lewis, C.E.; Workman, W.G.; Tumeo, M.A.; Hok, C.I.; Birklid, C.A.; Bennett, F.L.

    1988-12-01

    The ninety-ninth US Congress commissioned a six-state food irradiation research and development program to evaluate the commercial potential of this technology. Hawaii, Washington, Iowa, Oklahoma and Florida as well as Alaska have participated in the national program; various food products including fishery products, red meats, tropical and citrus fruits and vegetables have been studied. The purpose of the Alaskan study was to review and evaluate those factors related to the technical and economic feasibility of an irradiator in Alaska. This options analysis study will serve as a basis for determining the state's further involvement in the development of food irradiation technology. 40 refs., 50 figs., 53 tabs.

  15. One-dimensional migration of interstitial clusters in SUS316L and its model alloys at elevated temperatures

    NASA Astrophysics Data System (ADS)

    Satoh, Y.; Abe, H.; Matsukawa, Y.; Matsunaga, T.; Kano, S.; Arai, S.; Yamamoto, Y.; Tanaka, N.

    2015-05-01

    For self-interstitial atom (SIA) clusters in various concentrated alloys, one-dimensional (1D) migration is induced by electron irradiation around 300 K. But at elevated temperatures, the 1D migration frequency decreases to less than one-tenth of that around 300 K in iron-based bcc alloys. In this study, we examined mechanisms of 1D migration at elevated temperatures using in situ observation of SUS316L and its model alloys with high-voltage electron microscopy. First, for elevated temperatures, we examined the effects of annealing and short-term electron irradiation of SIA clusters on their subsequent 1D migration. In annealed SUS316L, 1D migration was suppressed and then recovered by prolonged irradiation at 300 K. In high-purity model alloy Fe-18Cr-13Ni, annealing or irradiation had no effect. Addition of carbon or oxygen to the model alloy suppressed 1D migration after annealing. Manganese and silicon did not suppress 1D migration after annealing but after short-term electron irradiation. The suppression was attributable to the pinning of SIA clusters by segregated solute elements, and the recovery was to the dissolution of the segregation by interatomic mixing under electron irradiation. Next, we examined 1D migration of SIA clusters in SUS316L under continuous electron irradiation at elevated temperatures. The 1D migration frequency at 673 K was proportional to the irradiation intensity. It was as high as half of that at 300 K. We proposed that 1D migration is controlled by the competition of two effects: induction of 1D migration by interatomic mixing and suppression by solute segregation.

  16. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    NASA Astrophysics Data System (ADS)

    Collette, R.; King, J.; Buesch, C.; Keiser, D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-07-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.

  17. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    DOE PAGES

    Collette, R.; King, J.; Buesch, C.; Keiser, Jr., D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-04-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

  18. Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

    SciTech Connect

    Gary S. Was

    2009-03-31

    The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.

  19. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    SciTech Connect

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed.

  20. The role of nickel in radiation damage of ferritic alloys

    DOE PAGES

    Osetskiy, Yury N.; Anento, Napoleon; Serra, Anna; Terentyev, Dmitry

    2014-11-26

    According to the modern theory damage evolution under neutron irradiation depends on the fraction of self interstitial atoms (SIAs) produced in the form of one-dimensionally (1-D) glissile clusters. These clusters, having a low interaction cross-section with other defects, sink mainly on grain boundaries and dislocations creating the so-called production bias. It is known empirically that addition of certain alloying elements affect many radiation effects, including swelling, however the mechanisms are unknown in many cases. In this paper we report the results of an extensive multi-technique atomistic level modeling of SIA clusters mobility in bcc Fe-Ni alloys with Ni content frommore » 0.8 to 10 at.%. We have found that Ni interacts strongly with periphery of clusters affecting their mobility. The total effect is defined by all Ni atoms interacting with the cluster at the same time and can be significant even in low-Ni alloys. Thus 1nm (37SIAs) cluster is practically immobile at T < 500K in the Fe-0.8at.% Ni alloy. Increasing cluster size and Ni content enhance cluster immobilization. Furthermore, this effect should have quite broad consequences in swelling rate, matrix damage accumulation, radiation induced hardening, etc. and the results obtained help in better understanding and prediction of radiation effects in Fe-Ni ferritic alloys.« less

  1. The role of nickel in radiation damage of ferritic alloys

    SciTech Connect

    Osetskiy, Yury N.; Anento, Napoleon; Serra, Anna; Terentyev, Dmitry

    2014-11-26

    According to the modern theory damage evolution under neutron irradiation depends on the fraction of self interstitial atoms (SIAs) produced in the form of one-dimensionally (1-D) glissile clusters. These clusters, having a low interaction cross-section with other defects, sink mainly on grain boundaries and dislocations creating the so-called production bias. It is known empirically that addition of certain alloying elements affect many radiation effects, including swelling, however the mechanisms are unknown in many cases. In this paper we report the results of an extensive multi-technique atomistic level modeling of SIA clusters mobility in bcc Fe-Ni alloys with Ni content from 0.8 to 10 at.%. We have found that Ni interacts strongly with periphery of clusters affecting their mobility. The total effect is defined by all Ni atoms interacting with the cluster at the same time and can be significant even in low-Ni alloys. Thus 1nm (37SIAs) cluster is practically immobile at T < 500K in the Fe-0.8at.% Ni alloy. Increasing cluster size and Ni content enhance cluster immobilization. Furthermore, this effect should have quite broad consequences in swelling rate, matrix damage accumulation, radiation induced hardening, etc. and the results obtained help in better understanding and prediction of radiation effects in Fe-Ni ferritic alloys.

  2. Correlation between diffusion barriers and alloying energy in binary alloys.

    PubMed

    Vej-Hansen, Ulrik Grønbjerg; Rossmeisl, Jan; Stephens, Ifan E L; Schiøtz, Jakob

    2016-01-28

    In this paper, we explore the notion that a negative alloying energy may act as a descriptor for long term stability of Pt-alloys as cathode catalysts in low temperature fuel cells. Using density functional theory calculations, we show that there is a correlation between the alloying energy of an alloy, and the diffusion barriers of the minority component. Alloys with a negative alloying energy may show improved long term stability, despite the fact that there is typically a greater thermodynamic driving force towards dissolution of the solute metal over alloying. In addition to Pt, we find that this trend also appears to hold for alloys based on Al and Pd. PMID:26750475

  3. Radiation Resistance Studies of Amorphous Silicon Alloy Photovoltaic Materials

    NASA Technical Reports Server (NTRS)

    Woodyard, James R.

    1994-01-01

    The radiation resistance of commercial solar cells fabricated from hydrogenated amorphous silicon alloys was investigated. A number of different device structures were irradiated with 1.0 MeV protons. The cells were insensitive to proton fluences below 1E12 sq cm. The parameters of the irradiated cells were restored with annealing at 200 C. The annealing time was dependent on proton fluence. Annealing devices for one hour restores cell parameters for fluences below lE14 sq cm require longer annealing times. A parametric fitting model was used to characterize current mechanisms observed in dark I-V measurements. The current mechanisms were explored with irradiation fluence, and voltage and light soaking times. The thermal generation current density and quality factor increased with proton fluence. Device simulation shows the degradation in cell characteristics may be explained by the reduction of the electric field in the intrinsic layer.

  4. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    SciTech Connect

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  5. Economics of food irradiation

    SciTech Connect

    Deitch, J.

    1982-01-01

    This article examines the cost competitiveness of the food irradiation process. An analysis of the principal factors--the product, physical plant, irradiation source, and financing--that impact on cost is made. Equations are developed and used to calculate the size of the source for planned product throughput, efficiency factors, power requirements, and operating costs of sources, radionuclides, and accelerators. Methods of financing and capital investment are discussed. A series of tables show cost breakdowns of sources, buildings, equipment, and essential support facilities for both a cobalt-60 and a 10-MeV electron accelerator facility. Additional tables present irradiation costs as functions of a number of parameters--power input, source size, dose, and hours of annual operation. The use of the numbers in the tables are explained by examples of calculations of the irradiation costs for disinfestation of grains and radicidation of feed.

  6. Effects of helium injection mode on void formation in Fe-Ni-Cr alloys

    NASA Astrophysics Data System (ADS)

    Kimoto, T.; Lee, E. H.; Mansur, L. K.

    1988-09-01

    The effect of the helium injection mode on void formation during ion irradiation of the pure solution-annealing alloys Fe-15Ni-7Cr, Fe-35Ni-7Cr, Fe-45Ni-7Cr, Fe-10Ni-13Cr, Fe-40Ni-13Cr, Fe-45Ni-15Cr was examined. Ion irradiation was carried out with 4 MeV Ni ions at 948 K to doses of 30 to 100 dpa with: (1) no helium injection, (2) simultaneous helium injection and (3) helium preinjection and aging. Swelling variation with helium injection differed among the 7Cr alloys and 13-15Cr alloys. Only the simultaneous helium injection mode produced a bimodal cavity size distribution in the high Ni alloys. The critical radius, as estimated from the cavity size distributions appears to have increased with increasing dose, but no clear variation of the critical radius with composition was observed. Helium preinjection and one-hour aging at 948 K formed helium bubbles along the residual dislocations, while subsequent Ni irradiation caused void formation along the dislocation lines. The calculated helium concentration deduced from observable helium bubbles was low compared with the injected helium concentration in the alloys containing higher Ni and lower Cr.

  7. Radiation effects on the microstructure of a 9Cr-ODS alloy.

    SciTech Connect

    Gan, J.; Allen, T. R.; Birtcher, R. C.; Shutthanandan, S.; Thevothasan, S.; Materials Science Division; INL; Univ. of Wisconsin at Madison; PNNL

    2008-01-01

    Oxide dispersion strengthened (ODS) steels are prime candidates for high-temperature, high-dose cladding in advanced nuclear reactors. When a 9Cr-ODS alloy was irradiated with 5 MeV nickel ions at temperatures of 500-700 C to doses up to 150 dpa, there was no significant change in the dislocation arrangement. For oxide particles, there is a small shrinkage in size and increase in density with increasing irradiation dose. This work confirms that oxide particles and the microstructure of the 9Cr-ODS show minimal changes under irradiation at temperatures up to 700 C and doses up to 150 dpa.

  8. Transmutation-induced embrittlement of V-Ti-Ni and V-Ni alloys in HFIR

    SciTech Connect

    Ohnuki, S.; Takahashi, H.; Garner, F.A.; Pawel, J.E.

    1996-04-01

    Vanadium, V-1Ni, V-10Ti and V-10Ti-1Ni (at %) were irradiated in HFIR to doses ranging from 18 to 30 dpa and temperatures between 300 and 600C. Since the irradiation was conducted in a highly thermalized neutron spectrum without shielding against thermal neutrons, significant levels of chromium (15-22%) were formed by transmutation. The addition of such large chromium levels strongly elevated the ductile to brittle transition temperature. At higher irradiation temperatures radiation-induced segregation of transmutant Cr and solute Ti at specimen surfaces leads to strong increases in the density of the alloy.

  9. Total lymphoid irradiation

    SciTech Connect

    Sutherland, D.E.; Ferguson, R.M.; Simmons, R.L.; Kim, T.H.; Slavin, S.; Najarian, J.S.

    1983-05-01

    Total lymphoid irradiation by itself can produce sufficient immunosuppression to prolong the survival of a variety of organ allografts in experimental animals. The degree of prolongation is dose-dependent and is limited by the toxicity that occurs with higher doses. Total lymphoid irradiation is more effective before transplantation than after, but when used after transplantation can be combined with pharmacologic immunosuppression to achieve a positive effect. In some animal models, total lymphoid irradiation induces an environment in which fully allogeneic bone marrow will engraft and induce permanent chimerism in the recipients who are then tolerant to organ allografts from the donor strain. If total lymphoid irradiation is ever to have clinical applicability on a large scale, it would seem that it would have to be under circumstances in which tolerance can be induced. However, in some animal models graft-versus-host disease occurs following bone marrow transplantation, and methods to obviate its occurrence probably will be needed if this approach is to be applied clinically. In recent years, patient and graft survival rates in renal allograft recipients treated with conventional immunosuppression have improved considerably, and thus the impetus to utilize total lymphoid irradiation for its immunosuppressive effect alone is less compelling. The future of total lymphoid irradiation probably lies in devising protocols in which maintenance immunosuppression can be eliminated, or nearly eliminated, altogether. Such protocols are effective in rodents. Whether they can be applied to clinical transplantation remains to be seen.

  10. Effects of neutron irradiation on strength of fusion reactor materials and their electron beam welded joints

    NASA Astrophysics Data System (ADS)

    Kaga, S.; Tamura, T.; Yoshida, H.; Miyata, K.

    1991-03-01

    Several aluminum alloys (A7N01, A5083 and A6061) and a ferritic martensitic steel (JFMS) were used in the present study of the effects of neutron irradiation on the strength of base materials and their electron beam welded joints. Neutron irradiation tests were performed using the core irradiation facility at Kyoto University Reactor (KUR). Neutron fluences were 2.0 × 10 22 9.1 × 10 22 and 1.7 × 10 23n/ m2 ( E > 0.1 MeV). Tensile tests were performed at 4.2, 77 and 293 K on miniature specimens prepared from both the base and welded materials. Aluminum alloys exhibit serrations in the nominal stress-nominal strain curve at 4.2 K. Little effect of neutron irradiation on the serration is observed. The ductility decrease of base metal and welded joints of aluminum alloys by neutron irradiation is smaller than that of JFMS. JFMS, especially welded joints, showed strong radiation embrittlement at cryogenic temperatures.

  11. Corrosion of the AlFeNi alloy used for the fuel cladding in the Jules Horowitz research reactor

    NASA Astrophysics Data System (ADS)

    Wintergerst, M.; Dacheux, N.; Datcharry, F.; Herms, E.; Kapusta, B.

    2009-09-01

    The AlFeNi aluminium alloy (1 wt% Fe, 1 wt% Ni, 1 wt% Mg) is expected to be used as nuclear fuel cladding for the Jules Horowitz experimental reactor. To guarantee a safe behaviour of the fuel, a good understanding of the fuel clad corrosion mechanisms is required. In this field, the experimental characterization of the selected alloy was performed. Then experimental studies of the aluminium alloy corrosion product obtained in autoclaves have shown an oxide film composed of two layers. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion parallel to a dissolution-precipitation process to form the outer zone. Dynamic experiments at 70 °C have demonstrated that a solid diffusion step controls the release kinetic. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core.

  12. Microstructure of V-4Cr-4Ti following low temperature neutron irradiation

    SciTech Connect

    Rice, P.M.; Snead, L.L.; Alexander, D.J.; Zinkle, S.J.

    1996-12-31

    The V-4Cr-4Ti alloys displays excellent mechanical properties, including a ductile-to-brittle transition temperature (DBTT) below - 200 C in the unirradiated conditions. Samples were fission neutron- irradiated in HFBR to a 0.4 dpa dose at 100-275 C. Mechanical tests showed significant irradiation hardening which increased with irradiation temperature. Charpy impact testing also showed a dramatic increase in DBTT on the order of 100 to 350 C. The mechanical property changes are correlated with preliminary results from TEM analysis of the defect microstructure resulting from the low-dose neutron irradiations. TEM of the irradiated material showed a nearly constant defect density of {approximately}1.6x10{sup 23}m{sup -3}, with an average defect diameter of slightly greater than 3 nm.

  13. Cascade effects on the irradiation stability of amorphous SiOC

    NASA Astrophysics Data System (ADS)

    Su, Qing; Cui, Bai; Kirk, Marquis A.; Nastasi, Michael

    2016-02-01

    We studied the heavy ion radiation tolerance of amorphous silicon oxycarbide (SiOC) alloys by in situ Kr ion irradiation within a transmission electron microscopy. The amorphous SiOC thin films were grown via co-sputtering from SiO2 and SiC targets on a surface-oxidized Si (100) substrate. These films were irradiated by 1 MeV Kr ions at both room temperature and 300 °C with damage levels up to 5 displacements per atom (dpa). TEM characterization shows no sign of crystallization, void formation or segregation in all irradiated samples. Our findings suggest that SiOC alloys are a class of promising radiation tolerant materials.

  14. Review of Advances in Development of Vanadium Alloys and MHD Insulator Coatings

    SciTech Connect

    Muroga, T.; Chen, J. M.; Chernov, V. M.; Fukumoto, Kenichi; Hoelzer, D. T.; Kurtz, Richard J.; Nagasaka, T.; Pint, Bruce A.; Satou, M.; Suzuki, Atsuyuki; Watanabe, Hideo

    2007-08-01

    In this paper, recent progress in the development of low activation vanadium alloys and MHD insulator coatings for Li-self cooled blanket is overviewed. The research progress in vanadium alloys is highlighted by technology of fabricating creep tubes, comparison of thermal creep in vacuum and Li, understanding on impurity transfer between vanadium alloys and Li and its impact on mechanical properties, behavior of hydrogen and hydrogen isotopes, low dose irradiation effects on weld joints and exploration for advanced vanadium alloys. Major remaining issues of vanadium alloys are thermal and irradiation creep, helium effects on high temperature mechanical properties and radiation effects on low temperature fracture properties. A new promising candidate of Er2O3, which showed good compatibility with Li, was identified for MHD insulator coating on vanadium alloys. The coating technology has made a significant progressed for the new candidate material. Recent efforts are being focused on multi-layer coating and in-situ coating. Tests in flowing lithium conditions with temperature gradient are necessary for quantitative examination of the performance.

  15. Proof-of-Principle Measurements on Unirradiated Zirconium Alloys

    SciTech Connect

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2007-04-18

    The ability to determine fuel assembly burnup has important non-proliferation implications since proliferation activities involve either irradiating fuel assemblies to a much lower level of burnup than is normal in commercial Light Water Reactor (LWR) practice, and/or irradiation of separate targets. Similarly, a method of determining burnup could be used to confirm declared operation for a reactor that is operating under IAEA safeguards. It is possible to determine fuel assembly burnup by measuring gamma radiation from specific fission products; however this technique is only useable after the fuel assembly has been out of the reactor for at least a year, and is not very useful after the assembly has been out of the reactor for 10 years or more. The use of isotope ratio measurements to measure the level of neutron exposure that material has received is well-known for graphite applications. The current project is an attempt to demonstrate that isotope ratio measurements can be performed on zirconium alloys used in LWR fuel assemblies. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers proof-of-principle measurements done on unirradiated zirconium alloys, these measurements show that: Titanium 48/Titanium 49 ratios can be measured in zirconium alloys using a Secondary Ionization Mass Spectrometer (SIMS) - enough Titanium was present in each of 6 samples tried to allow resolving the peaks associated with each isotope, and correction of interfering ions. The Ti 48/49 ratio measured in unirradiated zirconium alloy is, within a narrow error band, the same as that found in natural, unirradiated zirconium.

  16. Semiconductor alloys - Structural property engineering

    NASA Technical Reports Server (NTRS)

    Sher, A.; Van Schilfgaarde, M.; Berding, M.; Chen, A.-B.

    1987-01-01

    Semiconductor alloys have been used for years to tune band gaps and average bond lengths to specific applications. Other selection criteria for alloy composition, and a growth technique designed to modify their structural properties, are presently considered. The alloys Zn(1-y)Cd(y)Te and CdSe(y)Te(1-y) are treated as examples.

  17. De-alloyed platinum nanoparticles

    DOEpatents

    Strasser, Peter; Koh, Shirlaine; Mani, Prasanna; Ratndeep, Srivastava

    2011-08-09

    A method of producing de-alloyed nanoparticles. In an embodiment, the method comprises admixing metal precursors, freeze-drying, annealing, and de-alloying the nanoparticles in situ. Further, in an embodiment de-alloyed nanoparticle formed by the method, wherein the nanoparticle further comprises a core-shell arrangement. The nanoparticle is suitable for electrocatalytic processes and devices.

  18. PLUTONIUM-URANIUM-TITANIUM ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-07-28

    A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

  19. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  20. Irradiation and food processing.

    PubMed

    Sigurbjörnsson, B; Loaharanu, P

    1989-01-01

    After more than four decades of research and development, food irradiation has been demonstrated to be safe, effective and versatile as a process of food preservation, decontamination or disinfection. Its various applications cover: inhibition of sprouting of root crops; insect disinfestation of stored products, fresh and dried food; shelf-life extension of fresh fruits, vegetables, meat and fish; destruction of parasites and pathogenic micro-organisms in food of animal origin; decontamination of spices and food ingredients, etc. Such applications provide consumers with the increase in variety, volume and value of food. Although regulations on food irradiation in different countries are largely unharmonized, national authorities have shown increasing recognition and acceptance of this technology based on the Codex Standard for Irradiated Foods and its associated Code of Practice. Harmonization of national legislations represents an important prerequisite to international trade in irradiated food. Consumers at large are still not aware of the safety and benefits that food irradiation has to offer. Thus, national and international organizations, food industry, trade associations and consumer unions have important roles to play in introducing this technology based on its scientific values. Public acceptance of food irradiation may be slow at the beginning, but should increase at a faster rate in the foreseeable future when consumers are well informed of the safety and benefits of this technology in comparison with existing ones. Commercial applications of food irradiation has already started in 18 countries at present. The volume of food or ingredients treated on a commercial scale varies from country to country ranging from several tons of spices to hundreds of thousands of tons of grains per annum. With the increasing interest of national authorities and the food industry in applying the process, it is anticipated that some 25 countries will use some 55 commercial

  1. Micromechanical tests of ion irradiated materials: Atomistic simulations and experiments

    SciTech Connect

    Shin, C.; Jin, H. H.; Kwon, J.

    2012-07-01

    We investigated irradiation effects on Fe-Cr binary alloys by using a nano-indentation combined with a continuous stiffness measurement (CSM) technique. We modeled the nano-indentation test by using a finite element method. We could extract the intrinsic hardness and the yield stress of an irradiation hardened region by using a so-called inverse method. SiC micro-pillars of various sizes were fabricated by mask and inductively coupled plasma etching technique and compressed by using flat punch nano-indentation. Compressive fracture strength showed a clear specimen size effect. Brittle-to-Ductile transition at room temperature was observed as the specimen size decreases. The effect of irradiation on the fracture strength of SiC micro-pillars was evaluated by performing ion irradiation with Si ions. We have performed molecular dynamics simulations of nano-indentation and nano-pillar compression tests. Radiation effect was observed which is found to be due to the interaction of dislocations nucleated by spherical indenter with pre-existing radiation defects (voids). These atomistic simulations are expected to significantly contribute to the investigation of the fundamental deformation mechanism of small scale irradiated materials. (authors)

  2. ORNL irradiation creep facility

    SciTech Connect

    Reiley, T.C.; Auble, R.L.; Beckers, R.M.; Bloom, E.E.; Duncan, M.G.; Saltmarsh, M.J.; Shannon, R.H.

    1980-09-01

    A machine was developed at ORNL to measure the rates of elongation observed under irradiation in stressed materials. The source of radiation is a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). This choice allows experiments to be performed which simulate the effects of fast neutrons. A brief review of irradiation creep and experimental constraints associated with each measurement technique is given. Factors are presented which lead to the experimental choices made for the Irradiation Creep Facility (ICF). The ICF consists of a helium-filled chamber which houses a high-precision mechanical testing device. The specimen to be tested must be thermally stabilized with respect to the temperature fluctuations imposed by the particle beam which passes through the specimen. Electrical resistance of the specimen is the temperature control parameter chosen. Very high precision in length measurement and temperature control are required to detect the small elongation rates relevant to irradiation creep in the test periods available (approx. 1 day). The apparatus components and features required for the above are presented in some detail, along with the experimental procedures. The damage processes associated with light ions are discussed and displacement rates are calculated. Recent irradiation creep results are given, demonstrating the suitability of the apparatus for high resolution experiments. Also discussed is the suitability of the ICF for making high precision thermal creep measurements.

  3. Surface modification of high temperature iron alloys

    DOEpatents

    Park, Jong-Hee

    1995-01-01

    A method and article of manufacture of a coated iron based alloy. The method includes providing an iron based alloy substrate, depositing a silicon containing layer on the alloy surface while maintaining the alloy at a temperature of about 700.degree. C.-1200.degree. C. to diffuse silicon into the alloy surface and exposing the alloy surface to an ammonia atmosphere to form a silicon/oxygen/nitrogen containing protective layer on the iron based alloy.

  4. Surface modification of high temperature iron alloys

    DOEpatents

    Park, J.H.

    1995-06-06

    A method and article of manufacture of a coated iron based alloy are disclosed. The method includes providing an iron based alloy substrate, depositing a silicon containing layer on the alloy surface while maintaining the alloy at a temperature of about 700--1200 C to diffuse silicon into the alloy surface and exposing the alloy surface to an ammonia atmosphere to form a silicon/oxygen/nitrogen containing protective layer on the iron based alloy. 13 figs.

  5. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    SciTech Connect

    M.K. Meyer; J. Gan; J.-F. Jue; D.D. Keiser; E. Perez; A. Robinson; D.M. Wachs; N. Woolstenhulme; G.L. Hofman; Y.-S. Kim

    2014-04-01

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  6. Work function of binary alloys

    NASA Astrophysics Data System (ADS)

    Ishii, Ryusuke; Matsumura, Katsunori; Sakai, Akira; Sakata, Toyo

    2001-01-01

    By utilizing the field emission method, we have studied the composition dependence of work function in NiCu and PtRh alloys. In PtRh alloys, we find that the work function falls below the linear interpolation, in agreement with the experimental results on AgAu alloys [Fain and McDavid, Phys. Rev. B 9 (1974) 5099]. On the other hand, the work function of NiCu alloys is found to show little systematic deviation from the linear interpolation. The observed negative deviation in PtRh alloys is not compatible with a simple theoretical prediction based on the electronic density of states.

  7. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  8. Effect of boron on post irradiation tensile properties of reduced activation ferritic steel (F-82H) irradiated in HFIR

    SciTech Connect

    Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi; Pawel, J.E.

    1994-12-31

    Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.

  9. Dissimilar joint characteristics of SiC and WC-Co alloy by laser brazing

    NASA Astrophysics Data System (ADS)

    Nagatsuka, K.; Sechi, Y.; Nakata, K.

    2012-08-01

    SiC and WC-Co alloys were joined by laser brazing with an active braze metal. The braze metal based on eutectic Ag-Cu alloy with additional Ti as an active element ranging from 0 to 2.8 mass% was sandwiched by the SiC block and WC-Co alloy plate. The brazing was carried out by selective laser beam irradiation on the WC-Co alloy plate. The content of Ti in the braze metal was required to exceed 0.6 mass% in order to form a brazed joint with a measurable shear strength. The shear strength increased with increasing Ti content up to 2.3 mass%Ti and decreased with a higher content.

  10. Electron irradiation experiments in support of fusion materials development

    SciTech Connect

    Gelles, D.S. ); Ohnuki, S.; Takahashi, H. ); Matsui, H. ); Kohno, Y. )

    1991-11-01

    Microstructural evolution in response to 1 MeV irradiation has been investigated for three simple ferritic alloys, pure beryllium, pure vanadium, and two simple vanadium alloys over a range of temperatures and doses. Microstructural evolution in Fe-3, -9, and -18Cr ferritic alloys is found to consist of crenulated, faulted a<100> loops and circular, unfaulted a/2 <111> loops at low temperatures, but with only unfaulted loops at high temperatures. The complex dislocation evolution is attributed to sigma phase precipifaults arising from chromium segregation to point defect sinks. Beryllium is found to be resistant to electron damage; the only effect observed was enhanced dislocation mobility. Pure vanadium, V-5Fe, and V-1Ni microstructural response was complicated by precipitation on heating to 400{degrees}C and above, but dislocation evolution was investigated in the range of room temperature to 300{degrees}C and at 600{degrees}C. The three materials behaved similarly, except that pure vanadium showed more rapid dislocation evolution. This difference does not explain the enhanced swelling observed in vanadium alloys.

  11. Dissolution of ordered precipitates under ion irradiation

    SciTech Connect

    Camus, E.; Bourdeau, F.; Abromeit, C.; Wanderka, N.; Wollenberger, H.

    1995-09-01

    The stability of the ordered {gamma}{prime} precipitates under 300-keV Ni{sup +} irradiation was investigated between room temperature and 623 K. The two competing mechanisms of destabilization by cascade producing irradiation, i.e. disordering and dissolution of the {gamma}{prime} precipitates in Nimonic PE16 alloy, has been studied separately by electron microscopy and field-ion microscopy with atom probe. At high temperatures, the precipitates are stable. At intermediate temperatures, the precipitates dissolve by ballistic mixing into the matrix, but the interface is restored by the radiation-enhanced atomic jumps. The order in the precipitates remains stable. At low temperatures, the precipitates are dissolved by atomic mixing. The dissolution proceeds in a diffusional manner with a diffusion coefficient normalized by the displacement rate D/K = 0.75 nm{sup 2}dpa{sup {minus}1}. The precipitates become disordered by a fluence of 0.1 dpa, whereas precipitate dissolution needs much higher fluences.

  12. Alloyed coatings for dispersion strengthened alloys

    NASA Technical Reports Server (NTRS)

    Wermuth, F. R.; Stetson, A. R.

    1971-01-01

    Processing techniques were developed for applying several diffusion barriers to TD-Ni and TD-NiCr. Barrier coated specimens of both substrates were clad with Ni-Cr-Al and Fe-Cr-Al alloys and diffusion annealed in argon. Measurement of the aluminum distribution after annealing showed that, of the readily applicable diffusion barriers, a slurry applied tungsten barrier most effectively inhibited the diffusion of aluminum from the Ni-Cr-Al clad into the TD-alloy substrates. No barrier effectively limited interdiffusion of the Fe-Cr-Al clad with the substrates. A duplex process was then developed for applying Ni-Cr-Al coating compositions to the tungsten barrier coated substrates. A Ni-(16 to 32)Cr-3Si modifier was applied by slurry spraying and firing in vacuum, and was then aluminized by a fusion slurry process. Cyclic oxidation tests at 2300 F resulted in early coating failure due to inadequate edge coverage and areas of coating porosity. EMP analysis showed that oxidation had consumed 70 to 80 percent of the aluminum in the coating in less than 50 hours.

  13. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  14. Economics of food irradiation

    NASA Astrophysics Data System (ADS)

    Kunstadt, Peter; Eng, P.; Steeves, Colyn; Beaulieu, Daniel; Eng, P.

    1993-07-01

    The number of products being radiation processed worldwide is constantly increasing and today includes such diverse items as medical disposables, fruits and vegetables, spices, meats, seafoods and waste products. This range of products to be processed has resulted in a wide range of irradiator designs and capital and operating cost requirements. This paper discusses the economics of low dose food irradiation applications and the effects of various parameters on unit processing costs. It provides a model for calculating specific unit processing costs by correlating known capital costs with annual operating costs and annual throughputs. It is intended to provide the reader with a general knowledge of how unit processing costs are derived.

  15. Fuel or irradiation subassembly

    DOEpatents

    Seim, O.S.; Hutter, E.

    1975-12-23

    A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins.

  16. TEM investigation on the microstructural evolution of Hastelloy N induced by Ar⁺ ion irradiation.

    PubMed

    Liu, Min; Lu, Yanling; Liu, Renduo; Zhou, Xingtai

    2014-02-01

    Hastelloy N alloy has been selected as the primary structure material for molten salt reactor. In this article, Hastelloy N alloy samples were irradiated to different doses at room temperature using 300 keV Ar(+) ions. The microstructural evolution was investigated by transmission electron microscopy (TEM) and energy-dispersive spectroscopy (EDS). Black dot defects emerged in sample irradiated at low dose (0.4 displacement per atom (dpa)), and they grew up with irradiation doses (0.4-2 dpa). A high density of small dislocation loops (nano meters in size) were observed in the sample irradiated to 4 dpa. When the ion dose increased to 12 dpa, complicated structures with defects (including dislocation lines, larger loops and smaller black dots) were observed. Dislocation networks were detected from high-angle annular dark field (HAADF) images. Larger dislocation loops (size: 30-80 nm) were visible in the sample irradiated to 40 dpa. Irradiation with dose of 120 dpa led to the formation of face-centered cubic nanocrystallites with preferred orientations.

  17. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect

    Cook, David Howard; Freels, James D; Ilas, Germina; Jolly, Brian C; Miller, James Henry; Primm, Trent; Renfro, David G; Sease, John D; Pinkston, Daniel

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  18. Shape Memory Alloy Actuator

    NASA Technical Reports Server (NTRS)

    Baumbick, Robert J. (Inventor)

    2000-01-01

    The present invention discloses and teaches a unique, remote optically controlled micro actuator particularly suitable for aerospace vehicle applications wherein hot gas, or in the alternative optical energy, is employed as the medium by which shape memory alloy elements are activated. In gas turbine powered aircraft the source of the hot gas may be the turbine engine compressor or turbine sections.

  19. Quinary metallic glass alloys

    DOEpatents

    Lin, X.; Johnson, W.L.

    1998-04-07

    At least quinary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10{sup 3}K/s. Such alloys comprise zirconium and/or hafnium in the range of 45 to 65 atomic percent, titanium and/or niobium in the range of 4 to 7.5 atomic percent, and aluminum and/or zinc in the range of 5 to 15 atomic percent. The balance of the alloy compositions comprise copper, iron, and cobalt and/or nickel. The composition is constrained such that the atomic percentage of iron is less than 10 percent. Further, the ratio of copper to nickel and/or cobalt is in the range of from 1:2 to 2:1. The alloy composition formula is: (Zr,Hf){sub a}(Al,Zn){sub b}(Ti,Nb){sub c}(Cu{sub x}Fe{sub y}(Ni,Co){sub z}){sub d} wherein the constraints upon the formula are: a ranges from 45 to 65 atomic percent, b ranges from 5 to 15 atomic percent, c ranges from 4 to 7.5 atomic percent, d comprises the balance, d{hor_ellipsis}y is less than 10 atomic percent, and x/z ranges from 0.5 to 2.

  20. Quinary metallic glass alloys

    DOEpatents

    Lin, Xianghong; Johnson, William L.

    1998-01-01

    At least quinary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10.sup.3 K/s. Such alloys comprise zirconium and/or hafnium in the range of 45 to 65 atomic percent, titanium and/or niobium in the range of 4 to 7.5 atomic percent, and aluminum and/or zinc in the range of 5 to 15 atomic percent. The balance of the alloy compositions comprise copper, iron, and cobalt and/or nickel. The composition is constrained such that the atomic percentage of iron is less than 10 percent. Further, the ratio of copper to nickel and/or cobalt is in the range of from 1:2 to 2:1. The alloy composition formula is: (Zr,Hf).sub.a (Al,Zn).sub.b (Ti,Nb).sub.c (Cu.sub.x Fe.sub.y (Ni,Co).sub.z).sub.d wherein the constraints upon the formula are: a ranges from 45 to 65 atomic percent, b ranges from 5 to 15 atomic percent, c ranges from 4 to 7.5 atomic percent, d comprises the balance, d.multidot.y is less than 10 atomic percent, and x/z ranges from 0.5 to 2.