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Sample records for irradiated uranium-molybdenum alloy

  1. Hot rolling of thick uranium molybdenum alloys

    DOEpatents

    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.

  2. 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

  3. 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.

  4. Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum

    SciTech Connect

    Yeon Soo Kim; G. L. Hofman; A. B. Robinson; D. M. Wachs

    2013-10-01

    Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to [approximately]5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

  5. Qualification of uranium-molybdenum alloy fuel -- conclusions of an international workshop

    SciTech Connect

    Snelgrove, J. L.; Languilee, A.

    2000-02-14

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-MO alloy fuel at a workshop held at Argonne National Laboratory on January 17--18, 2000. Consensus was reached that the qualification plans of the US RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper.

  6. 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

  7. As-cast uranium-molybdenum based metallic fuel candidates and the effects of carbon addition

    NASA Astrophysics Data System (ADS)

    Blackwood, Van Stephen

    The objective of this research was to develop and recommend a metallic nuclear fuel candidate that lowered the onset temperature of gamma phase formation comparable or better than the uranium-10 wt. pct. molybdenum alloy, offered a solidus temperature as high or higher than uranium-10 wt. pct. zirconium (1250°C), and stabilized the fuel phase against interaction with iron and steel at least as much as uranium-10 wt. pct. zirconium stabilized the fuel phase. Two new as-cast alloy compositions were characterized to assess thermal equilibrium boundaries of the gamma phase field and the effect of carbon addition up to 0.22 wt. pct. The first system investigated was uranium- x wt. pct. M where x ranged between 5-20 wt. pct. M was held at a constant ratio of 50 wt. pct. molybdenum, 43 wt. pct. titanium, and 7 wt. pct. zirconium. The second system investigated was the uranium-molybdenum-tungsten system in the range 90 wt. pct. uranium - 10 wt. pct. molybdenum - 0 wt. pct. tungsten to 80 wt. pct. uranium - 10 wt. pct. molybdenum - 10 wt. pct. tungsten. The results showed that the solidus temperature increased with increased addition of M up to 12.5 wt. pct. for the uranium-M system. Alloy additions of titanium and zirconium were removed from uranium-molybdenum solid solution by carbide formation and segregation. The uranium-molybdenum-tungsten system solidus temperature increased to 1218°C at 2.5 wt. pct. with no significant change in temperature up to 5 wt. pct. tungsten suggesting the solubility limit of tungsten had been reached. Carbides were observed with surrounding areas enriched in both molybdenum and tungsten. The peak solidus temperatures for the alloy systems were roughly the same at 1226°C for the uranium-M system and 1218°C for the uranium-molybdenum-tungsten system. The uranium-molybdenum-tungsten system required less alloy addition to achieve similar solidus temperatures as the uranium-M system.

  8. 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

  9. 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

  10. PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

    SciTech Connect

    Pierce, R; Ann Visser, A; James Laurinat, J

    2007-10-15

    The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum.

  11. 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.

  12. 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.

  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. Surface engineering of low enriched uranium-molybdenum

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van den Berghe, S.; Detavernier, C.

    2013-09-01

    Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

  15. 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.

  16. SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET

    SciTech Connect

    Laurinat, J

    2007-05-31

    H-Canyon Engineering (HCE) is evaluating the feasibility of processing material from the Super Kukla Prompt Burst Reactor, which operated at the Nevada Test Site from 1964 to 1978. This material is comprised of 90 wt % uranium (U) (at approximately 20% 235U enrichment) alloyed with 10 wt % molybdenum (Mo). The objective is to dissolve the material in nitric acid (HNO{sub 3}) in the H-Canyon dissolvers and then to process the dissolved material through H-Canyon First and Second Cycle solvent extraction. The U product from Second Cycle will be sent to the highly enriched uranium (HEU) blend down program. In the blend down program, enriched U from the 1EU product stream will be blended with natural U at a ratio of 1 part enriched U per 3.5 parts natural U to meet a reactor fuel specification of 4.95% 235U before being shipped for use by the Tennessee Valley Authority (TVA) in its nuclear plants. The TVA specification calls for <200 mg Mo/g U (200 ppm). Since natural U has about 10 mg Mo/g U, the required purity of the 1EU product prior to blending is about 800 mg Mo/g U, allowing for uncertainties. HCE requested that the Savannah River National Laboratory (SRNL) define a flowsheet for the safe and efficient processing of the U-10Mo material. This report presents a computational model of the solvent extraction portion of the proposed flowsheet. The two main objectives of the computational model are to demonstrate that the Mo impurity requirement can be met and to show that the solvent feed rates in the proposed flowsheet, in particular to 1A and 1D Banks, are adequate to prevent refluxing of U and thereby ensure nuclear criticality safety. SASSE (Spreadsheet Algorithm for Stagewise Solvent Extraction), a Microsoft Excel spreadsheet that supports Argonne National Laboratory's proprietary AMUSE (Argonne Model for Universal Solvent Extraction) code, was selected to model the U/Mo separation flowsheet. SASSE spreadsheet models of H-Canyon First and Second Cycle solvent

  17. 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.

  18. Electroslag remelt processing of irradiated vanadium alloys

    NASA Astrophysics Data System (ADS)

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

    1996-10-01

    This paper describes experimental efforts to investigate the potential of a slag remelting process for reducing radioactivity of irradiated vanadium alloys used in a fusion power production facility. The experiment determined the removal characteristics of four surrogate transmutation isotopes significant to accident safety expected in a V5Ti5Cr alloy irradiated under fusion conditions (Ca, Y, to simulate Sc, Mn, and Ar). Removal of these isotopes could decrease the accident risk of reprocessing irradiated vanadium and reusing it in fusion reactors. An electroslag remelt furnace was used in the experiment to melt and react the constituents using a calcium fluoride slag. The process achieved 90% removal of calcium and over 99% removal of yttrium. Analyses indicate that 40% of the manganese has been removed. Argon analysis of the refined ingots indicates that 99% of the argon was removed from the vanadium alloy.

  19. 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.

  20. 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.

  1. Irradiation-induced microstructural changes in alloy X-750

    SciTech Connect

    Kenik, E.A.

    1997-04-01

    Alloy X-750 is a nickel base alloy that is often used in nuclear power systems for it`s excellent corrosion resistance and mechanical properties. The present study examines the microstructure and composition profiles in a heat of Alloy X-750 before and after neutron irradiation.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Irradiation performance of FFTF drivers using the D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-12-31

    Five test assemblies similar in design to the Fast Flux Test Facility driver fuel assembly , but employing the alloy D9 in place of stainless steel 316 for duct, cladding, and wire wrap compnents were irradiated to demonstrate the improved performance of the new design. Results of post-irradiation examinations are discussed.

  9. 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.

  10. 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.

  11. 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)

  12. Irradiation damage in multicomponent equimolar alloys and high entropy alloys (HEAs).

    PubMed

    Nagase, Takeshi; Rack, Philip D; Egami, Takeshi

    2014-11-01

    To maintain sustainable energy supply and improve the safety and efficiency of nuclear reactors, development of new and advanced nuclear materials with superior resistance to irradiation damage is necessary. Recently, a new generation of structural materials, termed as multicomponent equimolar alloys and/or high entropy alloys (HEAs), are being developed. These alloys consist of multicomponent elements for maximizing the compositional entropy, which stabilizes the solid solution phase. In this paper, preliminary studies on the irradiation damage in equimolar alloys and HEAs by High Voltage Electron Microscopy (HVEM) are reported [1-4]. (1) ZrHfNb equimolar alloys [1, 2]A multicomponent ZrHfNb alloy was prepared by a co-sputtering process using elemental Zr, Hf, and Nb targets using an AJA International ATC 2000-V system. A single-phase bcc solid solution was obtained in the ZrHfNb alloy with an approximately equiatomic ratio of its constituent elements. The irradiation-induced structural change in the ZrHfNb equimolar alloys with the bcc solid solution structure was investigated by HVEM using the Hitachi H-3000 installed at Osaka University. The polycrystalline bcc phase shows high phase stability against irradiation damage at 298 K; the bcc solid solution phase, whose grain size was about 20 nm, remained as a main constituent phase even after the severe irradiation damage that reached 10 dpa. (2) CoCrCuFeNi HEAs [3]A single-phase fcc solid solution was obtained in a CoCrCuFeNi alloy. The microstructure of the alloy depended on the preparation technique: a nanocrystalline CoCrCuFeNi alloy with an approximately equiatomic ratio of its constituent elements was obtained by a co-sputtering process with multi-targets, while polycrystalline structures were formed when the arc-melting method was used. Both nanocrystalline and polycrystalline structures showed high phase stability against fast electron irradiation at temperatures ranging from 298 K to 973 K; a fcc

  13. Irradiation Stability of Uranium Alloys at High Exposures

    SciTech Connect

    McDonell, W.R.

    2001-03-26

    Postirradiation examinations were begun of a series of unrestrained dilute uranium alloy specimens irradiated to exposures up to 13,000 MWD/T in NaK-containing stainless steel capsules. This test, part of a program of development of uranium metal fuels for desalination and power reactors sponsored by the Division of Reactor Development and Technology, has the objective of defining the temperature and exposure limits of swelling resistance of the alloyed uranium. This paper discusses those test results.

  14. 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.

  15. 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.

  16. Effects of self-irradiation in plutonium alloys

    DOE PAGESBeta

    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.

  17. 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.

  18. 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.

  19. Neutron irradiation creep in stainless steel alloys

    NASA Astrophysics Data System (ADS)

    Schüle, Wolfgang; Hausen, Hermann

    1994-09-01

    Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300°C and 500°C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of "primary" creep stage is observed for doses up to 3-5 dpa after which dose the "secondary" creep stage begins. The "primary" creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These "primary" creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of α-ferrite below about 400°C and of carbides below about 700°C, and not to irradiation creep. The "secondary" creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature ( Qirr = 0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels.

  20. 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.

  1. 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.

  2. Irradiation-Induced Nanoprecipitation in Ni-W Alloys

    NASA Astrophysics Data System (ADS)

    Lee, Jaeyel; Lear, Calvin R.; Zhang, Xuan; Bellon, Pascal; Averback, Robert S.

    2015-03-01

    The evolution of Ni-W alloy thin films subjected to Kr irradiation at room temperature and subsequent annealing at 1123 K (850 °C) was studied by X-ray diffraction and transmission electron microscopy. Irradiation resulted in significant increase in grain size, from ~20 nm in the as-grown state to over 300 nm after irradiation and annealing. The compositions selected for the study, 18 and 23 at. pct W, resulted in the formation of an ordered Ni4W matrix after annealing. Remarkably, in the Ni-23 at. pct W films, irradiation followed by annealing induced the precipitation of two families of Ni2W4C carbides, large blocky ones at grain boundaries, and intragranular nanocarbides, ~5 to 20 nm in size and with a high number density, 9.0 × 1022 m-3. In contrast, only blocky Ni6W6C carbides formed in control specimens directly subjected to annealing. The intragranular Ni2W4C nanocarbides displayed an orientation relationship with the Ni4W matrix, and they appear to be effective traps for implanted Kr ions, since nanobubbles formed on their periphery. The results suggest that non-equilibrium processing can be used to nucleate nanocarbides in the grain interiors of Ni-W alloys, and that this may improve alloy properties, including radiation resistance.

  3. 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.

  4. Effects of self-irradiation in plutonium alloys

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

    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.

  5. 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.

  6. Microstructural development of neutron irradiated W?Re alloys

    NASA Astrophysics Data System (ADS)

    Nemoto, Yoshiyuki; Hasegawa, Akira; Satou, Manabu; Abe, Katsunori

    2000-12-01

    Tungsten (W) alloys are candidate materials to be used as high-heat-flux materials in fusion reactors. In our previous work, W-26 wt% Re showed drastic hardening and embrittlement after the neutron irradiation. In this study, to clarify the irradiation hardening and embrittlement behavior of W-26 wt% Re, from the viewpoint of microstructural development, the microstructure observation of the neutron irradiated W-26 wt% Re was carried out using transmission electron microscope (TEM). The specimens were irradiated at the materials open test assembly of the fast flux test facility (FFTF/MOTA-2A cycle 11) up to ˜1×10 27 n/m2, ( En>0.1 MeV). The irradiation temperatures were 646, 679, 792, 873 and 1073 K. In all neutron irradiated W-26 wt% Re samples, sigma-phase precipitates and chi-phase precipitates were observed, while in the thermally aged specimen, only sigma-phase precipitates were observed. Irradiation effects on microstructural development are discussed.

  7. 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.

  8. 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.

  9. 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.

  10. Radiation hardening of V C, V O, V N alloys neutron-irradiated to high fluences

    NASA Astrophysics Data System (ADS)

    Chuto, Toshinori; Satou, Manabu; Abe, Katsunori

    1998-10-01

    Vanadium has a large affinity for interstitial impurities such as C, N and O. Mechanical properties and irradiation performance of vanadium alloys are affected by the impurities. Radiation hardening and defect microstructures of vanadium alloys doped with relatively large amounts of these interstitial elements were studied. Neutron irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-1F) to 47.9 dpa at temperatures of 679, 793 and 873 K. Irradiation hardening decreased with increasing irradiation temperature. Increase in hardness for the V-C alloy was relatively greater after irradiation at the low temperatures. Decorated dislocations and voids were observed depending on the alloying elements. The factors for irradiation hardening were different for each interstitial element in the alloys irradiated at 873 K to 47.9 dpa.

  11. 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.

  12. 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)

  13. 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.

  14. Irradiation creep of various ferritic alloys irradiated at ˜400°C in the PFR and FFTF reactors

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    Irradiation creep of three ferritic alloys at ˜400 ∘C has been studied. Specimens were in the form of pressurized tubes. In a joint US/UK creep study, two identical sets of creep specimens constructed from one heat of HT9 were irradiated in fast reactors, one in the Prototypic Fast Reactor (PFR) and the other in the Fast Flux Test Facility (FFTF). The specimens in PFR were irradiated to a dose of ˜50 dpa, whereas the specimens in FFTF were irradiated to a dose of 165 dpa. The observed swelling and creep behavior were very different in the two reactors. Creep specimens constructed from D57, a developmental alloy ferritic alloy, were also irradiated in PFR to a dose of ˜50 dpa. Creep behavior typical of previous studies on ferritic alloys was observed. Finally, creep specimens constructed from MA957, a Y 2O 3 dispersion-hardened ferritic alloy, were irradiated in FFTF to a dose of ˜110 dpa. This alloy exhibited a large amount of densification, and the creep behavior was different than observed in more conventional ferritic or ferritic-martensitic alloys.

  15. Swelling in several commercial alloys irradiated to very high neutron fluence

    SciTech Connect

    Gelles, D.S.; Pintler, J.S.

    1983-01-01

    Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650/sup 0/C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Post-irradiation mechanical properties of an AlMgSi alloy

    NASA Astrophysics Data System (ADS)

    Ismail, Z. H.; Birt, B.

    1995-03-01

    The effect of fast-neutron irradiation on the tensile properties and hardness of the age-hardenable alloy AlMgSi is investigated. Post-irradiation tensile tests are carried out in the temperature range 298 to 628 K. The results show that the degree of irradiation-produced hardening is dependent upon the initial condition of the alloy. The alloy in its soft condition exhibits a higher degree of irradiation hardening compared with that in the hard condition. The implication of the results is discussed in terms of the variation in the microstructures involved and compared with previosly published data.

  2. 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.

  3. Neutron irradiation effects on the microstructure of low-activation ferritic alloys*1

    NASA Astrophysics Data System (ADS)

    Kimura, A.; Matsui, H.

    1994-09-01

    Microstructures of low-activation ferritic alloys, such as 2.25% Cr-2% W, 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys, were observed after FFTF irradiation at 698 K to a dose of 36 dpa. Martensite in 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys and bainite in 2.25% Cr-2% W alloy were fairly stable after the irradiation. Microvoids were observed in the martensite in each alloy but not in bainite and δ-ferrite in 12% Cr-2% W alloys. An addition of 0.02% Ti to 9% Cr-2% W alloy considerably reduced the void density. Spherical (Ta, W) and Ti-rich precipitates were observed in the Ti-added 9% Cr-2% W alloy. Precipitates observed in 9% Cr-2% W and 7% Cr-2% W alloys are mainly Cr-rich M 23C 6 (Ta, W) and Ta(W)-rich M 6C and Fe-rich Laves phase. In 2.25% Cr-2% W alloy, high density of fine (Ta, W)-rich M 2C type precipitates as well as M 6C were observed. Spherical small α' Cr-rich particles were observed in both martensite and α-ferrite in 12% Cr-2% W alloys. Correlation between postirradiation microstructure and irradiation hardening is shown and discussed for these alloys.

  4. 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.

  5. 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.

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

    DOE PAGESBeta

    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

  7. 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.

  8. 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.

  9. 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.

  10. Microstructures of deformed VTiCrSi type alloys after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Satou, Manabu; Abe, Katsunori; Kayano, Hideo

    1996-10-01

    The alloy of V5Ti5Cr1SiAl,Y (nominal composition, weight percentage) was developed to improve oxidation properties and high temperature strength, and has been studied as one of the candidates for fusion applications. This alloy showed low swelling properties and enough tensile ductility after neutron irradiation to high fluence levels. The dislocation microstructures after tensile deformation and defect microstructures in the neutron-irradiated alloy to high fluences were studied. Irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-2A) at 406°C to 46 dpa and the deformation microstructures were examined by transmission electron microscopy. Slip dislocations were developed inhomogeneously in the specimen deformed at ambient temperature after neutron irradiation. Dislocation loops contributed mainly to hardening of the alloy after irradiation; however, cavities and radiation-induced precipitates did not so much.

  11. 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.

  12. Heavy ion irradiation induced dislocation loops in AREVA's M5® alloy

    NASA Astrophysics Data System (ADS)

    Hengstler-Eger, R. M.; Baldo, P.; Beck, L.; Dorner, J.; Ertl, K.; Hoffmann, P. B.; Hugenschmidt, C.; Kirk, M. A.; Petry, W.; Pikart, P.; Rempel, A.

    2012-04-01

    Pressurized water reactor (PWR) Zr-based alloy structural materials show creep and growth under neutron irradiation as a consequence of the irradiation induced microstructural changes in the alloy. A better scientific understanding of these microstructural processes can improve simulation programs for structural component deformation and simplify the development of advanced deformation resistant alloys. As in-pile irradiation leads to high material activation and requires long irradiation times, the objective of this work was to study whether ion irradiation is an applicable method to simulate typical PWR neutron damage in Zr-based alloys, with AREVA's M5® alloy as reference material. The irradiated specimens were studied by electron backscatter diffraction (EBSD), positron Doppler broadening spectroscopy (DBS) and in situ transmission electron microscopy (TEM) at different dose levels and temperatures. The irradiation induced microstructure consisted of - and -type dislocation loops with their characteristics corresponding to typical neutron damage in Zr-based alloys; it can thus be concluded that heavy ion irradiation under the chosen conditions is an excellent method to simulate PWR neutron damage.

  13. 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.

  14. 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.

  15. Separation of radiation defects in Ni and Ni-C alloys under electron and neutron irradiation

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

    Complex investigations of radiation damage of Ni and Ni- 880 at. ppm C alloy under electron and neutron irradiation in the region of room temperature hardened and deformed state. In pure nickel, with the deformation microstructure, both in electron and in the neutron irradiation is observed separation of radiation-induced defects. When electron irradiation in the alloy Ni-C separation effect is observed, and when neutron irradiation there is no. This is due to the interaction of carbon atoms with radiation defects. The main sinks for radiation-induced defects are the areas with a high concentration of defects in cascades of atomic displacements.

  16. Radiation-induced softening of Fe-Mo alloy under high- temperature electron irradiation

    NASA Astrophysics Data System (ADS)

    Tsepelev, A.

    2016-04-01

    Effect of radiation-induced change of mechanical properties of Fe-5 wt.% Mo alloy irradiated with electrons (2 MeV) at room temperature and 400°C has been investigated. Mechanical properties were estimated by Miniaturized Disk Bend Test technique. Effect of radiation softening of the alloy is ascertained the value of which was increased with temperature rise. With the purpose of separation of thermal and radiation contributions into the effects, the tests were carried out for specimens annealed in the same thermal conditions (temperature and duration of annealing) just as during irradiation. Thermal annealing and electron irradiation at 400°C is found to bring to multidirectional effects of the alloy strengthening and softening respectively. It is concluded that irradiation suppresses the effect of thermal-induced strengthening and stimulates a softening of the alloy due to more significant changes in the structure and phase composition of it.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. Recombination mechanism of point defect loss to coherent precipitates in alloys under irradiation

    NASA Astrophysics Data System (ADS)

    Turkin, A. A.; Bakai, A. S.

    A new mechanism of defect loss by enhanced recombination inside coherent precipitates in alloys under irradiation is described. The mechanism is examined quantitatively to find the microstructural parameters responsible for resistance to dimensional instability. The proposed model explains why radiation properties of Zr-Nb alloys depend on density of fine-grained precipitates of β Nb-phase.

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

    DOE PAGESBeta

    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.

  3. 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

  4. Tensile behavior and microstructure of neutron-irradiated Mo-5% Re alloy

    NASA Astrophysics Data System (ADS)

    Hasegawa, Akira; Abe, Katsunori; Satou, Manabu; Namba, Chusei

    1995-08-01

    This work reports the effect of heat treatment on the tensile behavior and and microstructure of neutron-irradiated Mo-5% Re alloy. Stress-relived and recrystallized specimens conditions were irradiated at five temperatures between 646 and 1073 K in FFTF/MOTA. The exposure levels were in the range of 6.8 to 34 dpa depending on the irradiation temperatures. Tensile tests were carried out at room temperature and 673 K and microstructures of the irradiated specimens were observed by TEM. The Mo-5% Re alloy irradiated at high temperatures shows ductile behavior even at room temperature. The total elongation of stress-relived specimens irradiated at 873 and 1073 K ranged from 5 to 10%, and that of recrystallized specimens irradiated at 1073 K was 5%. The fracture modes of these specimens were transgranular type. Voids were observed in all of the irradiated specimens, but precipitates were found only in specimens irradiated above 792 K. It is important for the Mo sbnd Re alloy to be used in high-heat flux components of fusion reactors that the alloy showed the ductility after neutron exposures of relatively high fluences.

  5. 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.

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

    DOE PAGESBeta

    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

  7. 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.

  8. 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.

  9. 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.

  10. Potential Annealing Treatments for Tailoring the Starting Microstructure of Low-Enriched U-Mo Dispersion Fuels to Optimize Performance During Irradiation

    SciTech Connect

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Nicolas E. Woolstenhulme; Ashley Ewh

    2011-12-01

    Low-enriched uranium-molybdenum alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 C for up to 4 hours. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt% Si and a heat treatment temperature of at least 475 C.

  11. 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.

  12. 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

  13. Helium generation rates in isotopically tailored Fe-Cr-Ni alloys irradiated in FFTF/MOTA

    SciTech Connect

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

    1991-11-01

    Three Fe-Cr-Ni alloys have been doped with 0.4% {sup 59}Ni for side-by-side irradiations of doped and undoped materials in order to determine the effects of fusion-relevant levels of helium production on microstructural development and mechanical properties. The alloys were irradiated in three successive cycles of the Materials Open Test Assembly (MOTA) located in the Fast Flux Test Facility (FFTF). Following irradiation, helium levels were measured by isotope dilution mass spectrometry. The highest level of helium achieved in doped alloys was 172 appm at 9.1 dpa for a helium(appm)-to-dpa ratio of 18.9. The overall pattern of predicted helium generation rates in doped and undoped alloys is in good agreement with the helium measurements.

  14. AFIP-6 Irradiation Summary Report

    SciTech Connect

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

    2011-09-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 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.

  15. Neutron irradiation embrittlement of molybdenum rhenium alloys and their improvement by heat treatment

    NASA Astrophysics Data System (ADS)

    Hasegawa, Akira; Ueda, Kazukiyo; Satou, Manabu; Abe, Katsunori

    1998-10-01

    Irradiation-induced embrittlement is one of the major problems with molybdenum alloys, which have been considered as one of the candidates for divertor structural materials. The effects of rhenium content, heat-treatment and irradiation condition on mechanical properties and microstructural development of molybdenum rhenium alloys were studied after neutron exposure by FFTF/MOTA up to high fluence (˜1 × 10 27 n/m 2, En>0.1 MeV). Appreciable plastic deformation was observed in a bending test on stress-relieved Mo-5 wt% Re irradiated at high temperatures. Fine, dense precipitates were observed in Mo-41 wt% Re, resulting in large hardening and embrittlement.

  16. 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.

  17. 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.

  18. 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.

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

    DOE PAGESBeta

    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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Mechanical behavior and microstructural evolution of vanadium alloys irradiated in ATR-A1

    NASA Astrophysics Data System (ADS)

    Fukumoto, K.-i.; Matsui, H.; Tsai, H.; Smith, D. L.

    2000-12-01

    An irradiation experiment has been done in the ATR-A1 to investigate irradiation behavior of vanadium alloys in the low temperature regime from 200°C to 300°C with damage levels of 3 to 4 dpa. In creep measurements, creep tubes of V-3Fe-4Ti-0.1Si with inner pressures up to 165 MPa did not rupture during irradiation. The effective strain rate of creep was below 0.2% dpa-1 and it showed the same tendency as V-4Cr-4Ti alloys. In Charpy impact tests, all specimens of V-4Cr-4Ti-0.1Si and V-3Fe-4Ti-0.1Si showed brittle behavior at room temperature and the DBTT increased to 60-150°C. The fracture surface showed cleavage. Tensile tests conducted both at room temperature and at the irradiation temperature showed significant irradiation hardening and brittle responses. TEM showed that high densities of tiny defect clusters were formed in V-Cr-Ti and V-Fe-Ti alloys. Precipitates could not be seen in specimens irradiated below 300°C, however, fine defect clusters are considered to be the origin of brittle behavior in V-Cr-Ti alloys irradiated at low temperatures.

  7. 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.

  8. A phase field model for segregation and precipitation induced by irradiation in alloys

    NASA Astrophysics Data System (ADS)

    Badillo, A.; Bellon, P.; Averback, R. S.

    2015-04-01

    A phase field model is introduced to model the evolution of multicomponent alloys under irradiation, including radiation-induced segregation and precipitation. The thermodynamic and kinetic components of this model are derived using a mean-field model. The mobility coefficient and the contribution of chemical heterogeneity to free energy are rescaled by the cell size used in the phase field model, yielding microstructural evolutions that are independent of the cell size. A new treatment is proposed for point defect clusters, using a mixed discrete-continuous approach to capture the stochastic character of defect cluster production in displacement cascades, while retaining the efficient modeling of the fate of these clusters using diffusion equations. The model is tested on unary and binary alloy systems using two-dimensional simulations. In a unary system, the evolution of point defects under irradiation is studied in the presence of defect clusters, either pre-existing ones or those created by irradiation, and compared with rate theory calculations. Binary alloys with zero and positive heats of mixing are then studied to investigate the effect of point defect clustering on radiation-induced segregation and precipitation in undersaturated solid solutions. Lastly, irradiation conditions and alloy parameters leading to irradiation-induced homogeneous precipitation are investigated. The results are discussed in the context of experimental results reported for Ni-Si and Al-Zn undersaturated solid solutions subjected to irradiation.

  9. 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.

  10. 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.

  11. Swelling and dislocation evolution in simple ferritic alloys irradiated to high fluence in FFTF/MOTA

    NASA Astrophysics Data System (ADS)

    Katoh, Yutai; Kohyama, Akira; Gelles, David S.

    1995-08-01

    Microstructures of a series of Fe sbnd Cr binary ferritic alloys were examined following neutron irradiation to 140 dpa at 698 K in FFTF/MOTA. The chromium concentration ranged from 3 to 18% in 3% increments and the irradiation temperature corresponded to the peak swelling condition for this alloy class. The swelling varied from 0.4 to 2.9% depending on chromium concentration, and the highest swelling was found in the Fe sbnd 9Cr alloy. The cavity microstructures corresponded to transient to early steady-state swelling regime. Dislocations were composed of networks with both a<100> and ( a/2)<111> Burgers vector and a<100> type interstitial loops. The dislocation density was negatively correlated with swelling. Explanation for the observed chromium concentration dependence of microstructural development and low swelling in the ferritic alloys will be studied in connection with the dislocation bias efficiency and the theory of sink strength ratio.

  12. 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.

  13. Thermal properties of U-Mo alloys irradiated to moderate burnup and power

    NASA Astrophysics Data System (ADS)

    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 3.30 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm-3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  14. 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.

  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

    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

    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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

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

    DOE PAGESBeta

    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.

  3. ac susceptibility of thermally annealed and neutron irradiated Cu-Ni alloys

    NASA Technical Reports Server (NTRS)

    Catchings, R. M., III; Borg, R. J.; Violet, C. E.

    1985-01-01

    Thermal annealing and high-flux neutron irradiation are used to vary the degree of short-range atomic order in Cu-Ni alloys of composition 40, 50, and 60 at. pct Ni. The magnetic state is measured by ac magnetic susceptibility measurements. It is shown that annealing at 350 C causes significant changes in the susceptibility of all the samples. In the 50 and 60 at. pct Ni samples, the transition is broadened and extended to higher temperatures, while the 40 at. pct Ni sample changes from a paramagnetic system to a weakly ferromagnetic system. The neutron irradiation, in contrast to the thermal treatment, causes the development of smaller size cluster formations. The irradiated 60 at. pct Ni sample exhibits no change in the shape of its susceptibility curve from that of the quenched sample, whereas, the 40 pct alloy is changed, by irradiation, from a paramagnetic system to a spin-glass system.

  4. A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

    NASA Astrophysics Data System (ADS)

    Onimus, Fabien; Béchade, Jean-Luc

    2009-02-01

    Zirconium alloys used as fuel cladding tubes in the nuclear industry undergo important changes after neutron irradiation in the microstructure as well as in the mechanical properties. However, the effects of the specific post-irradiation deformation mechanisms on the mechanical behavior are not clearly understood and modeled. Based on experimental results it is discussed that the kinematic strain hardening is increased by the plastic strain localization inside the dislocation channels as well as the only basal slip activation observed for specific mechanical tests. From this analysis, the first polycrystalline model is developed for irradiated zirconium alloys, taking into account the irradiation induced hardening, the intra-granular softening as well as the intra-granular kinematic strain hardening due to the plastic strain localization inside the channels. This physically based model reproduces the mechanical behavior in agreement with the slip systems observed. In addition, this model reproduces the Bauschinger effect observed during low cycle fatigue as well as the cyclic strain softening.

  5. 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.

  6. Swelling in commercial Fe-Cr-Ni based alloys under electron irradiation

    NASA Astrophysics Data System (ADS)

    Thomas, L. E.; Gelles, D. S.

    1982-08-01

    Electron irradiation in a 1 MeV electron microscope has been used to study the void swelling response of several commercial austenitic stainless steels and iron-nickel based superalloys. Use of the 1 MeV microscope permits direct, continuous observation of the void development during elevated-temperature irradiations at displacement rates about 10 000 times greater then those in a fast breeder reactor. The alloys examined in this work included AISI 310, RA 330, A286, M813, Nimonic PE16, Inconel 706, Inconel 718 and Incoloy 901. Both helium preinjected specimens and uninjected specimens were studied. In all of the above alloys, swelling proceeds by formation of irradiation-induced dislocations and voids, followed by growth of the voids. The swelling rates and peak swelling temperatures vary considerably with alloy composition, heat treatment and helium preinjection. Comparisons of these results with recently reported swelling data from the same alloys after high fluence neutron irradiation in the EBR-II reactor shows good qualitative agreement in most cases. Helium preinjection of the electron irradiated specimens generally produced a poorer simulation than no helium preinjection. In one or two cases where the electron and neutron irradiation results strongly disagree, the differences appear to result from differences in irradiation-induced precipitation. Although the correlations between neutron and electron irradiation results are inadequate to obtain reliable engineering data by simulation, in-reactor swelling behavior is in general qualitatively well-represented by swelling response in the 1 MeV electron microscope. Nimonic is the registered trademark of Henry Wiggin and Company, UK. Inconel and Incoloy are registered trademarks of the International Nickel Company, Inc.

  7. Nanoindentation creep study on an ion beam irradiated oxide dispersion strengthened alloy

    NASA Astrophysics Data System (ADS)

    Huang, Zijing; Harris, Adrian; Maloy, Stuart A.; Hosemann, Peter

    2014-08-01

    Oxide dispersion strengthened (ODS) alloys are considered advanced structural materials for nuclear application due to their radiation tolerance and creep resistance. Ion beam irradiation is used to study the property changes due to displacement damage. In this work 1 dpa displacement damage in an ODS was produced followed by a nanoindentation creep study at temperatures up to 600 °C to evaluate the changes in mechanical properties due to irradiation. Converted yield strength (YS) and creep related parameters are reported.

  8. 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.

  9. 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.

  10. The influence of microstructure on blistering and bubble formation by He ion irradiation in Al alloys

    NASA Astrophysics Data System (ADS)

    Soria, S. R.; Tolley, A.; Sánchez, E. A.

    2015-12-01

    The influence of microstructure and composition on the effects of ion irradiation in Al alloys was studied combining Atomic Force Microscopy, Scanning Electron Microscopy and Transmission Electron Microscopy. For this purpose, irradiation experiments with 20 keV He+ ions at room temperature were carried out in Al, an Al-4Cu (wt%) supersaturated solid solution, and an Al-5.6Cu-0.5Si-0.5Ge (wt.%) alloy with a very high density of precipitates, and the results were compared. In Al and Al-4Cu, He bubbles were found with an average size in between 1 nm and 2 nm that was independent of fluence. The critical fluence for bubble formation was higher in Al-4Cu than in Al. He bubbles were also observed below the critical fluence after post irradiation annealing in Al-4Cu. The incoherent interfaces between the equilibrium θ phase and the Al matrix were found to be favorable sites for the formation of He bubbles. Instead, no bubbles were observed in the precipitate rich Al-5.6Cu-0.5Si-0.5Ge alloy. In all alloys, blistering was observed, leading to surface erosion by exfoliation. The blistering effects were more severe in the Al-5.6Cu-0.5Si-0.5Ge alloy, and they were enhanced by increasing the fluence rate.

  11. 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.

  12. Simulation of the nanostructure evolution under irradiation in Fe-C alloys

    NASA Astrophysics Data System (ADS)

    Jansson, V.; Malerba, L.

    2013-11-01

    Neutron irradiation induces in steels nanostructural changes, which are at the origin of the mechanical degradation that these materials experience during operation in nuclear power plants. Some of these effects can be studied by using as model alloy the iron-carbon system.

  13. Alloying effect of Ni and Cr on irradiated microstructural evolution of type 304 stainless steels

    NASA Astrophysics Data System (ADS)

    Tan, L.; Busby, J. T.

    2013-11-01

    Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ˜320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

  14. 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.

  15. 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.

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

    DOE PAGESBeta

    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

  17. Response of unirradiated and neutron-irradiated vanadium alloys to Charpy-impact loading*1

    NASA Astrophysics Data System (ADS)

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

    1991-03-01

    The ductile-brittle transition temperature (DBTT) was determined by Charpy-impact tests for dehydrogenated ( < 30 appm H) and hydrogenated (400-1200 appm H) V-7.2Cr-14.5Ti, V-9.9Cr-9.2Ti, V-13.5Cr-5.2Ti, V-17.7Ti, V-9.2Cr-4.9Ti, V-9.0Cr-3.2Fe-1.2Zr, V-3.1Ti-0.5Si, V-4.1Cr-4.3Ti, V-4.6Ti, and V-2.5Ti-1.0Si alloys. The DBTT was also determined for V-13.5Cr-5.2Ti, V-9.2Cr-4.9Ti, V-7.2Cr-14.5Ti, and V-17.7Ti alloys after neutron irradiation at 420 and 600°C to 41-44 dpa. The DBTTs determined for these vanadium alloys show that a vanadium alloy containing Cr and/or Ti and Si alloying additions to be used as a structural material in a fusion reactor should contain 3-9 wt% total alloying addition for maximum resistance to hydrogen- and/or irradiation-induced embrittlement.

  18. 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.

  19. Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation.

    PubMed

    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

  20. 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.

  1. 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

  2. Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

    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.

  3. 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.

  4. Resistance of (Fe, Ni)/sub 3/V long-range-ordered alloys to neutron and ion irradiation

    SciTech Connect

    Braski, D.N.

    1981-01-01

    A series of (Fe, Ni)/sub 3/V long-range-ordered alloys were irradiated with neutrons in the Oak Ridge Research Reactor (ORR) and with 4 MeV Ni ions at temperatures above 250/sup 0/C. The displacement damage levels for the two irradiations were 3.8 and 70 dpa, and the helium levels were 29 and 560 at. ppM, respectively. Irradiation in ORR generally increased the yield strength and lowered the ductility of an LRO alloy, but produced relatively little swelling. The LRO alloys retained their long-range order after ion irradiation below the critical ordering temperature, T/sub c/, and exhibited low swelling. Above T/sub c/ the alloys were disordered and showed greater swelling. Adjustment of alloy composition to prevent sigma phase formation reduced swelling.

  5. 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}.

  6. Microstructural changes in dilute Ni-Be alloys during HVEM sub-threshold irradiations

    SciTech Connect

    Regnier, P.G.; Lam, N.Q.

    1984-10-01

    The microstructural sensitivity of Ni-0.7 at.% Be alloys to HVEM sub-threshold irradiations was investigated. Several aspects were examined: (1) dose dependence of the microstructure change during irradiation with 350-keV electrons at 350/sup 0/C; (2) energy dependence of the incubation dose for the first appearance of black dots in the alloy films under irradiation at 350/sup 0/C along various crystallographic directions; and (3) temperature dependence of the microstructural evolution during 350-keV electron irradiation. It was found that sub-threshold irradiations were capable of inducing nonequilibrium solute segregation. Below approx. 400/sup 0/C, segregation-induced homogeneous precipitation of the ..gamma..-phase occurred in the alloy matrix, whereas at higher temperatures, only heterogeneous precipitation was observed at defect sinks. From the information about the energy dependence of the incubation dose for precipitation, the displacement threshold energy for Ni and point-defect production rate by secondary Be-Ni collisions were estimated.

  7. 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.

  8. 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.

  9. 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.

  10. Comparison of irradiation creep and swelling of an austenitic alloy irradiated in FFTF and PFR

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.; Munro, B.; Adaway, S.; Standring, J.

    1999-10-01

    comparative irradiation of identically constructed creep tubes in the Fast Flux Test Facility (FFTF) and the Prototypic Fast Reactor (PFR) shows that differences in irradiation conditions arising from both reactor operation and the design of the irradiation vehicle can have a significant impact on the void swelling and irradiation creep of austenitic stainless steels. In spite of these differences, the derived creep coefficients fall within the range of previously observed values for 316 SS.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2007-01-01

    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 (˜700 °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.

  13. V-alloy embrittlement by irradiation in a cooling gas environment

    NASA Astrophysics Data System (ADS)

    Röhrig, H. D.; Rieth, M.; Dafferner, B.; Materna-Morris, E.

    2000-12-01

    Sub-sized tensile and Charpy specimens of the vanadium-alloy V-4Cr-4Ti have been irradiated in a piggy-back experiment with ferritic-martensitic material in the HFR at Petten at 250-450°C to about 2.5 dpa, and subsequently tested. A complex set of post-irradiation mechanical properties was found. Impact measurements showed severe shifts in ductile-brittle transition temperatures (DBTT) regardless of irradiation temperature and no clear trends of recovery at higher irradiation temperatures. The fracture surfaces always indicated transcrystalline crack propagation. Tensile tests generally revealed irradiation hardening with a maximum around 300°C, accompanied by total elongations which increased with increasing temperature. Extremely small uniform elongations at the lower temperatures suggest high localized plastic deformation under small strain rates. Deviations from expected irradiation effects led to a scrutiny of the atmospheric conditions during irradiation which revealed that the capsule had several times been flushed with He and Ne containing traces of interstitial gases or N2 for temperature adjustment. A skin was found on the specimens irradiated at the highest temperatures. X-ray diffraction results showed an uptake of nitrogen, whereas microprobe analysis suggested the presence of oxygen in the bulk. The limited data allow only a qualitative discrimination between chemical and irradiation effects.

  14. Tensile and impact properties of vanadium-base alloys irradiated at <430°C

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    Tensile and impact properties were investigated at <430°C on V-Cr-Ti, V-Ti-Si, and V-Ti alloys after irradiation to ≈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 ≈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 ≈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.

  15. 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.

  16. 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.

  17. Evaluation of radiation hardening in ion-irradiated Fe based alloys by nanoindentation

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Ren, Ai; Jiang, Jing; Xu, Chaoliang; Huang, Ping; Qian, Wangjie; Wu, Yichu; Zhang, Chonghong

    2014-01-01

    Nanoindentation in combination with ion irradiation offers the possibility to quantify irradiation hardening due to radiation damage. Irradiation experiments for Fe-1.0wt.%Cu alloys, China A508-3 steels, and 16MND5 steels were carried out at about 100 °C by proton and Fe-ions with the energy of 240 keV, 3 MeV respectively. The constant stiffness measurement (CSM) with a diamond Berkovich indenter was used to obtain the depth profile of hardness. The results showed that under 240 keV proton irradiation (peak damage up to 0.5 dpa), Fe-1.0wt.%Cu alloys exhibited the largest hardening (∼55%), 16MND5 steels resided in medium hardening (∼46%), and China A508-3(2) steels had the least hardening (∼10%). Under 3 MeV Fe ions irradiation (peak damage up to 1.37 dpa), both China A508-3(1) and 16MND5 steels showed the same hardening (∼26%). The sequence of irradiation tolerance for these materials is China A508-3(2) > 16MND5 ≈ China A508-3(1) > Fe-1.0wt.%Cu. Based on the determination of the transition depth, the nominal hardness H0irr was also calculated by Kasada method.

  18. 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).

  19. Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Almazouzi, A.

    2009-03-01

    The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.

  20. 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.

  1. 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.

  2. A laboratory method to predict hydriding properties of zirconium alloys under irradiation

    SciTech Connect

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

    1988-08-01

    A corrosion and hydriding test series on zirconium alloys in the Engineering Test Reactor G-7 loop demonstrated relatively large lot-to-lot and alloy-to-alloy differences in hydriding rates under irradiation. Similar differences were also found among irradiated Zircaloy-2 pressure tubes fabricated by three suppliers for the Hanford Site N Reactor. This substantial in-reactor hydriding data base and access to archive materials from these alloys permitted an investigation of methods to reproduce the in-reactor hydriding orders-of-merit by an out-of-reactor method. The out-of-reactor method selected for investigation consisted of autoclaving alloys in relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solutions. The test times ranged from 7 to 35 d, and the samples were held at constant temperatures within the water reactor coolant temperature range (280/degree/C to 315/degree/C). The in-reactor hydriding behavior for several lots of Zircaloy-2, one lot of Zircaloy-4, and one lot of Zr--2.5Nb was reproduced in the lithium hydroxide tests. The hydriding rates were compared on the basis of the ratio of hydrogen weight gain to oxide weight gain. 13 refs., 4 figs., 6 tabs.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Power ultrasound irradiation during the alkaline etching process of the 2024 aluminum alloy

    NASA Astrophysics Data System (ADS)

    Moutarlier, V.; Viennet, R.; Rolet, J.; Gigandet, M. P.; Hihn, J. Y.

    2015-11-01

    Prior to any surface treatment on an aluminum alloy, a surface preparation is necessary. This commonly consists in performing an alkaline etching followed by acid deoxidizing. In this work, the use of power ultrasound irradiation during the etching step on the 2024 aluminum alloy was studied. The etching rate was estimated by weight loss, and the alkaline film formed during the etching step was characterized by glow discharge optical emission spectrometry (GDOES) and scanning electron microscope (SEM). The benefit of power ultrasound during the etching step was confirmed by pitting potential measurement in NaCl solution after a post-treatment (anodizing).

  9. 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.

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. Correlating radiation exposure with embrittlement: Comparative studies of electron- and neutron-irradiated pressure vessel alloys

    SciTech Connect

    Alexander, D. E.; Rehn, L. E.; Odette, G. R.; Lucas, G. E.; Klingensmith, D.; Gragg, D.

    1999-12-22

    Comparative experiments using high energy (10 MeV) electrons and test reactor neutrons have been undertaken to understand the role that primary damage state has on hardening (embrittlement) induced by irradiation at 300 C. Electrons produce displacement damage primarily by low energy atomic recoils, while fast neutrons produce displacements from considerably higher energy recoils. Comparison of changes resulting from neutron irradiation, in which nascent point defect clusters can form in dense cascades, with electron irradiation, where cascade formation is minimized, can provide insight into the role that the in-cascade point defect clusters have on the mechanisms of embrittlement. Tensile property changes induced by 10 MeV electrons or test reactor neutron irradiations of unalloyed iron and an Fe-O.9 wt.% Cu-1.0 wt.% Mn alloy were examined in the damage range of 9.0 x 10{sup {minus}5} dpa to 1.5 x 10{sup {minus}2} dpa. The results show the ternary alloy experienced substantially greater embrittlement in both the electron and neutron irradiate samples relative to unalloyed iron. Despite their disparate nature of defect production similar embrittlement trends with increasing radiation damage were observed for electrons and neutrons in both the ternary and unalloyed iron.

  15. Ultrasonic irradiation and its application for improving the corrosion resistance of phosphate coatings on aluminum alloys.

    PubMed

    Sheng, Minqi; Wang, Chao; Zhong, Qingdong; Wei, Yinyin; Wang, Yi

    2010-01-01

    In this paper, ultrasonic irradiation was utilized for improving the corrosion resistance of phosphate coatings on aluminum alloys. The chemical composition and morphology of the coatings were analyzed by X-ray diffraction analysis (XRD) and scanning electron microscopy (SEM). The effect of ultrasonic irradiation on the corrosion resistance of phosphate coatings was investigated by polarization curves and electrochemical impedance spectroscopy (EIS). Various effects of the addition of Nd(2)O(3) in phosphating bath on the performance of the coatings were also investigated. Results show that the composition of phosphate coating were Zn(3)(PO(4))(2).4H(2)O(hopeite) and Zn crystals. The phosphate coatings became denser with fewer microscopic holes by utilizing ultrasonic irradiation treatment. The addition of Nd(2)O(3) reduced the crystallinity of the coatings, with the additional result that the crystallites were increasingly nubby and spherical. The corrosion resistance of the coatings was also significantly improved by ultrasonic irradiation treatment; both the anodic and cathodic processes of corrosion taking place on the aluminum alloy substrate were suppressed consequently. In addition, the electrochemical impedance of the coatings was also increased by utilizing ultrasonic irradiation treatment compared with traditional treatment. PMID:19692286

  16. Correlation of microstructure and tensile and swelling behavior of neutron-irradiated vanadium alloys

    SciTech Connect

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

    1991-10-01

    The microstructures of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were characterized by transmission electron microscopy (TEM) after neutron irradiation in the Fast Flux Test Facility (FFTF) at 420 and 600{degrees}C to influences up to 114 dpa. Two types of irradiation-induced precipitates were identified, i.e., Ti{sub 2}O and Ti{sub 5}(Si,P){sub 3}. Blocky Ti(O,N,C) precipitates, which form by thermal processes during ingot fabrication, also were observed in all unirradiated and irradiated specimens. Irradiation-induced precipitation of spherical (<15 nm in diameter) Ti{sub 5}(Si,P){sub 3} phase was associated with superior resistance to void swelling. In specimens with negligible swelling, Ti{sub 5}(Si,P){sub 3} precipitation was significant. It seems that ductility is significantly reduced when the precipitation of Ti{sub 2}O and Ti{sub 5}(Si,P){sub 3} is pronounced. These observations indicate that initial composition; fabrication processes; actual solute compositions of Ti, O, N, C, P, and Si after fabrication; O, N, and C uptake during service; and irradiation-induced precipitation ae interrelated and are important factors to consider in developing an optimized alloy. 15 refs., 8 figs.

  17. 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.

  18. 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.

  19. Structural and magnetic properties of He+ irradiated Co2MnSi Heusler alloys

    NASA Astrophysics Data System (ADS)

    Abdallah, I.; Ratel-Ramond, N.; Magen, C.; Pecassou, B.; Cours, R.; Arnoult, A.; Respaud, M.; Bobo, J. F.; BenAssayag, G.; Snoeck, E.; Biziere, N.

    2016-04-01

    We have investigated the atomic disorder induced by a 150 keV He+ ion irradiation in a 40 nm thick Co2MnSi Heusler alloy. Disorder parameters on each atomic site are deduced from normal and anomalous x-ray diffraction measurements with Co and Cu Kα sources. While the film grows mainly in the L21 phase with inclusion of B2 grains, we observe an increase of both the Mn–Si and Co–Mn exchanges with the ion fluence. HAADF-STEM analysis demonstrates that the increase in Mn–Si disorder corresponds to a growing size of the B2 grains while the Co–Mn exchange is accounted for a D03 disorder type in the L21 matrix. These structural modifications are shown to decrease the average magnetization of the alloy, which is due to D03 disorder and local defects induced by irradiation.

  20. 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.

  1. 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.

  2. Monte Carlo simulations of copper clustering in Fe-Cu alloys under irradiation

    NASA Astrophysics Data System (ADS)

    Kwon, J.; Kwon, S. C.; Hong, J. H.

    2004-10-01

    We present the computational approach for studying the microstructures of Cu clusters in Fe-Cu alloys by combining the molecular dynamics (MD) simulation and Monte Carlo methods. The MD simulation is used to characterize the primary damage resulting from the displacement cascade in Fe. Then, using the Metropolis Monte Carlo methods, the microstructure of the Cu clusters is predicted under the assumption that the system will evolve towards the equilibrium state. The formation of the Cu clusters is apparent for Fe-Cu alloys of a higher Cu content (1.0 w/o), whereas the degree of Cu clustering is not significant for the lower Cu content (0.1 w/o) alloys. The atomic configuration of the Cu-vacancy complex under irradiation, produced by this simulation, is in a fair agreement with the experiments. The simulation is expected to provide important information on the Cu-cluster morphology, which is useful for experimental data analysis.

  3. 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.

  4. Defect structure development in a pure iron and dilute iron alloys irradiated with neutrons and electrons

    NASA Astrophysics Data System (ADS)

    Okada, A.; Maeda, H.; Hamada, K.; Ishida, I.

    The defect structure and mechanical property changes were observed for pure iron of 99.99% purity and a series of Fe-(0.1% and 0.4%) Cr and Fe-(0.1% and 0.4%) Mn dilute alloys irradiated with neutrons. From the comparison of the defect structures with yield strength change, a large contribution of the invisible defect clusters to the irradiation hardening was expected in the specimens irradiated in Japan Materials Test Reactor (JMTR), whereas these clusters are not found after irradiation in the Fast Flux Test Facility (FFTF). The electron irradiation experiments showed that addition of chromium and manganese to 0.1% in pure iron, development of large dislocation loops is suppressed, and frequent nucleation of small loops at the early stage of the electron irradiation is observed, similar to that in ultra-high purity iron of 99.9999% purity. The mechanisms of dislocation loop development in the early stage of irradiation for Fe-Cr and Fe-Mn are considered to be different.

  5. 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.

  6. Solute segregation and microstructural evolution in ion-irradiated vanadium-base alloys

    SciTech Connect

    Loomis, B.A.; Kestel, B.J.; Gerber, S.B.

    1986-02-01

    The microstructural evolution in V+14.7 a/o Cr, V+2.5 a/o Mo, V+2.5 a/o W, V+8.0 a/o Ni, V+5.3 a/o Ti, V+14.7 a/o Cr + 5.2 a/o Ti, V+9.6 a/o Cr + 3.1 a/o Fe+0.7 a/o Zr, and V+3.2 a/o Ti + 1.8 a/o Si alloys, and unalloyed V was determined from observations of specimens by TEM following 4.0-MeV /sup 51/V/sup + +/ ion irradiation at 900 to 970/sup 0/K to 50 dpa. The RIS of solutes in the ion-irradiated alloys was investigated by AES, EDXS, and EELS analyses. The RIS of solutes and microstructural evolution in the irradiated alloys were correlated with the solute diffusivity and the relative chemical affinity of the substitutional solutes for oxygen.

  7. Release of deuterium from irradiation damage in Fe-9Cr-2W ferritic alloy irradiated with deuterium ions

    NASA Astrophysics Data System (ADS)

    Ono, K.; Miyamoto, M.; Kudo, F.

    2014-09-01

    The release profile of deuterium from an Fe-9Cr-2W ferritic alloy irradiated with low-energy deuterium ions was studied by thermal desorption spectroscopy (TDS) and in situ transmission electron microscopy (TEM). It was found that one sharp TDS peak appeared at a temperature around 410 K depending on the heating rate that ranged from 1.5 to 20 K/min. The TDS peak height increased with increasing fluence from 2 × 1019 to 2 × 1021 D+/m2 with no shift of the peak temperature. A close correlation between these TDS peaks and the disappearance of dislocation loops formed by the irradiation was observed. The effects of tiny bubbles on TDS were small. These results suggest that most of the deuterium was trapped by dislocation loops, which affected the thermal stability of dislocation loops in the alloy. The dependence of TDS peak temperature on the heating rate yielded an activation energy of 0.63 ± 0.02 eV for deuterium de-trapping from dislocation loops. The retention properties of the total amount of deuterium exhibited a tendency of saturation at values on the order of 1020 D+/m2, which corresponded to a saturation tendency of the loop density.

  8. 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.

  9. 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.

  10. Microstructural changes in a neutron-irradiated Fe-15 at.%Cr alloy

    NASA Astrophysics Data System (ADS)

    Bachhav, Mukesh; Robert Odette, G.; Marquis, Emmanuelle A.

    2014-11-01

    Microstructural changes in a Fe-15 at.%Cr model alloy neutron irradiated to 1.82 dpa at 290 °C were characterized by atom probe tomography. Homogenously distributed α‧ precipitates as well as fewer clusters containing Si, P, Ni, and Cr, were observed in the matrix. Grain boundary analyses before and after irradiation revealed segregation of Cr, with W-shape concentration profiles developing in the vicinity of grain boundary carbide and nitride particles. After irradiation, impurities such as C, Si and P were segregated to the grain boundaries. Zones depleted of α‧ clusters, and Si were found at the interfaces of carbide and nitride precipitates and along grain boundaries in the vicinity of these precipitates.

  11. Auger electron spectroscopy study of alloy 718 and 304L stainless steel irradiated with 800 MeV protons

    NASA Astrophysics Data System (ADS)

    García-Mazarío, M.; Hernández-Mayoral, M.; Lancha, A. M.

    2001-07-01

    It is well known that radiation produces changes in materials microstructure such as formation of defects, dissolution and redistribution of secondary phases, precipitation of new phases, etc. and changes in the grain boundary microchemistry by a process known as radiation-induced segregation (RIS). This paper describes the grain boundary microchemical characterization of alloy 718 and 304L stainless steel irradiated with high-energy protons at Los Alamos Neutron Science Center (LANSCE), performed by means of Auger electron spectroscopy (AES). In addition, non-irradiated alloy 718 was characterized as reference. The Auger results showed that as a consequence of exposure to proton radiation, the changes observed in alloy 718 were the disappearance of the nickel and niobium rich grain boundaries precipitates and RIS of the major alloying elements (nickel to grain boundaries, and chromium and iron away from grain boundaries). On the other hand, in irradiated AISI 304L no differences were observed between intergranular and transgranular areas.

  12. Effects of neutron irradiation and hydrogen on ductile-brittle transition temperatures of V-Cr-Ti alloys*1

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

    The effects of neutron irradiation and hydrogen on the ductile-brittle transition temperatures (DBTTs) of unalloyed vanadium and V-Cr-Ti alloys were determined from Charpy-impact tests on {1}/{3} ASTM-standard-size specimens and from impact tests on 3-mm diameter discs. The tests were conducted on specimens containing < 30 appm hydrogen and 600-1200 appm hydrogen and on specimens after neutron irradiation to 28-46 atom displacements per atom at 420, 520, and 600°C. The DBTTs were minimum (< -220°C) for V-(1-5)Ti alloys and for V-4Cr- 4Ti alloy with < 30 appm hydrogen. The effect of 600-1200 appm hydrogen in the specimens was to raise the DBTTs by 60-100°C. The DBTTs were minimum (< -200°C) for V-(3-5)Ti and V-4Cr-4Ti alloys after neutron irradiation.

  13. 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.

  14. Spectroscopic monitoring on irradiation-induced formation of AuAg alloy nanoparticles by femtosecond laser

    NASA Astrophysics Data System (ADS)

    Herbani, Yuliati; Nakamura, Takahiro; Sato, Shunichi

    2016-02-01

    The interaction of an intense femtosecond laser pulses with a neat liquid solvent has been known to produce a number of highly reactive species that are useful to induce chemical reactions in the solution through the nonlinear absorption processes. When metal ions are present in the solution, they are assumed to readily reduce by ions, radicals, molecules or excited states generated photolytically from the solvent resulting in the formation of zerovalent metal nanoalloys. If two kinds of metal precursors are involved in a reduction process, the alloying process is expected. In this work, irradiation-induced synthesis of AuAg bimetallic nanoparticle at different laser-pulse energies was examined to investigate the formation mechanism in the presence of NH4OH in the initial solution. At a given laser pulse energy (5.8 mJ/pulse), the time evolution of the UV-visible absorption spectra showed that the formation of AuAg nanoalloys most likely begin with the formation of Ag-riched alloy nanoparticles. As the reduction is started, the absorption spectrum of solution was closer to that of pure Ag nanoparticles. This indicates that the reduction rate of Ag is relatively greater than that of Au nanoparticle in the presence of ammonia. The single peak position then shifts to the red region as the irradiation time increases. After 10 min, the peak positions are between pure silver and gold peaks indicating the alloying process occurs at this stage. At low pulse energy (1.0 mJ/pulse), there was an induction time for several minutes before the absorption is detectable, and hence the alloying process is also delayed (after 20 minutes irradiation). While the formation rate of nanoparticles is more pronounced at high laser pulse energy, the formation yield is relatively the same for both laser pulse energies.

  15. Surface modifications of hydrogen storage alloy by heavy ion beams with keV to MeV irradiation energies

    NASA Astrophysics Data System (ADS)

    Abe, Hiroshi; Tokuhira, Shinnosuke; Uchida, Hirohisa; Ohshima, Takeshi

    2015-12-01

    This study deals with the effect of surface modifications induced from keV to MeV heavy ion beams on the initial reaction rate of a hydrogen storage alloy (AB5) in electrochemical process. The rare earth based alloys like this sample alloy are widely used as a negative electrode of Ni-MH (Nickel-Metal Hydride) battery. We aimed to improve the initial reaction rate of hydrogen absorption by effective induction of defects such as vacancies, dislocations, micro-cracks or by addition of atoms into the surface region of the metal alloys. Since defective layer near the surface can easily be oxidized, the conductive oxide layer is formed on the sample surface by O+ beams irradiation, and the conductive oxide layer might cause the improvement of initial reaction rate of hydriding. This paper demonstrates an effective surface treatment of heavy ion irradiation, which induces catalytic activities of rare earth oxides in the alloy surface.

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

    SciTech Connect

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

    2004-01-31

    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. (nanoparticles)

  17. Properties of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

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

    1996-02-01

    One property of vanadium-base alloys that is not well understood in terms of their potential use as fusion reactor structural materials is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of {approx} 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18--31 dpa at 425--600 C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V-5Ti, V-3Ti-1Si, V-8Cr-6Ti, and V-4Cr-4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420 C. However, postirradiation ductilities at < 250 C were higher than those of the non-DHCE specimens (< 0.1 appm helium), whereas strengths were lower, indicating that different types of hardening centers are produced during DHCE and non-DHCE irradiation. Ductile-brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > {minus}150 C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at {minus}196 C in some specimens that were irradiated to 31 dpa at 425 C during the DHCE. For the helium generation rates in this experiment ({approx} 0.4--4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

  18. 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.

  19. Microstructure and tensile properties of heavily irradiated 5052-0 aluminum alloy

    SciTech Connect

    Farrell, K.

    1980-01-01

    During neutron irradiation of an aluminum 2.2% magnesium solid solution alloy in the High Flux Isotope Reactor to fast and thermal fluences > 10/sup 27/ neutrons (n)m/sup 2/ at 328/sup 0/K (0.35 T/sub m/) about seven percent insoluble, transmutant silicon was produced. Some of this silicon reacted with the dissolved magnesium to form a fine precipitate of Mg/sub 2/Si. A tight dislocation structure was also created. The alloy showed good resistance to cavity formation. Tension tests at 323, 373, and 423/sup 0/K (0.35, 0.40, and 0.45 T/sub m/) showed pronounced irradiation-induced strengthening and an associated marked loss in ductility. These changes were greater than in magnesium-free aluminum and in alloys containing preexisting, thermally-aged Mg/sub 2/Si precipitate. Increasing the thermal-to-fast flux ratio from 1.7 to 2.1 caused further strengthening beyond that expected from a simple increase in silicon level.

  20. 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.

  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. Microstructural development in reduced activation ferritic alloys irradiated to 200 dpa at 420$deg;C

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1994-09-01

    Density change and microstructural development are reported for nine reduced activation ferritic steels covering the range 2.3 to 12Cr with varying additions of V and/or W for hardening and up to 6.5 Mn for austenite stability. Specimens were examined following irradiation in FFTF/MOTA at 420°C to a dose exceeding 200 dpa. Void swelling was found, but the swelling remained at 5% or below, with the worst case in an alloy of 9Cr-2Mn-1WV. The carbide structure pinning martensite lath boundaries remained in place.

  3. Silicon's role in determining swelling in neutron-irradiated Fe-Cr-Ni-Si alloys

    SciTech Connect

    Sekimura, N. ); Garner, F. A. ); Newkirk, J.W. )

    1991-11-01

    Two silicon-modified alloy series, one based on Fe-15Cr-20Ni and another based on Fe-15Cr-25Ni were irradiated at target temperatures between 399 and 649{degree}C in EBR-II. The influence of silicon on swelling is more complex than previously envisioned and indicates that silicon plays two or more competing roles while in solution. Radiation-induced formation of {gamma}{prime} (Ni{sub 3}Si) precipitates is dependent on silicon and nickel content, as well as temperature. Precipitation of {gamma}{prime} appears to play only a minor role in void formation.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Surface composition, microstructure and corrosion resistance of AZ31 magnesium alloy irradiated by high-intensity pulsed ion beam

    SciTech Connect

    Li, P.; Lei, M.K.; Zhu, X.P.

    2011-06-15

    High-intensity pulsed ion beam (HIPIB) irradiation of AZ31 magnesium alloy is performed and electrochemical corrosion experiment of irradiated samples is carried out by using potentiodynamic polarization technology in order to explore the effect of HIPIB irradiation on corrosion resistance of magnesium alloy. The surface composition, cross-sectional morphology and microstructure are characterized by using electron probe microanalyzer, optical microscope and transmission electron microscope, respectively. The results indicated that HIPIB irradiation leads to a significant improvement in corrosion resistance of magnesium alloy, in terms of the considerable increase in both corrosion potential and pitting breakdown potential. The microstructural refinement and surface purification induced by HIPIB irradiation are responsible for the improved corrosion resistance. - Research Highlights: {yields} A modified layer about 30 {mu}m thick is obtained by HIPIB irradiation. {yields} Selective ablation of element/impurity phase having lower melting point is observed. {yields} More importantly, microstructural refinement occurred on the irradiated surface. {yields} The modified layer exhibited a significantly improved corrosion resistance. {yields} Improved corrosion resistance is ascribed to the combined effect induced by HIPIB.

  9. Positron Annihilation Study of Zr-2.5 wt.% Nb alloy Irradiated by Ar9+ heavy ions

    NASA Astrophysics Data System (ADS)

    Devi, Aruna; Menon, Ranjini; Maheshwari, Priya; Neogy, S.; Mukherjee, P.; Nabhiraj, P. Y.; Pujari, P. K.; Srivastava, D.; Dey, G. K.

    2015-06-01

    Zr-2.5 Nballoy is used as a pressure tube material in pressurized heavy water reactor (PHWR). It is one of the most critical component which decides the lifespan of the reactor. The in-reactor degrading phenomenon of prime concern is dimensional changes caused by irradiation induced creep and growth processes. The present study aims to understand the mechanism of irradiation damage by irradiating the alloy with heavy ion. Such type of irradiation study would facilitate larger damage of material in a shorter time. Zr-2.5Nb alloy samples were irradiated using 315 keV Ar9+ ion for different durations. The irradiation doses were varied in the range of 3.1X1015 to 4.17X1016 Ar9+/cm2. SRIM calculation was carried out to evaluate damage profile in the irradiated samples. Beam based Positron Annihilation Spectroscopy (PAS) technique was used for depth profiling to characterize defect distribution in the alloys. The no. of defects generated is seen to increase with the increase in the fluence.

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

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-06-01

    Six test assemblies similar in design to the FFTF driver assembly but employing the advanced alloy D9 in place of Type 316 stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of this design. A single pinhole-type breach was incurred in one of the high exposure tests after a peak fuel burnup of 155 MWd/kgM and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared to analytical evaluations. A revised swelling correlation for D9 Alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup greater than 120 MWd/kgM. The extended lifetime capability of this design would reduce fuel supply requirements for the FFTF by a third relative to the reference driver design.

  15. Low void swelling in dispersion strengthened copper alloys under single-ion irradiation

    NASA Astrophysics Data System (ADS)

    Hatakeyama, M.; Watanabe, H.; Akiba, M.; Yoshida, N.

    2002-12-01

    Oxide dispersion strengthened copper (ODS-Cu) alloys GlidCop CuAl15 and CuAl25 were irradiated with Cu 2+ ions at 573-773 K up to doses of 30 dpa. Void swelling was observed in all specimens irradiated at temperatures ranging from 573 to 673 K. In CuAl15 brazed with graphite at 1083 K, mean grain size was about 800 nm. Voids were observed in grains larger than 1 μm but not in smaller than 500 nm in diameter. The CuAl25 joined with SUS316 by hot isostatic pressing (HIP) at 1323 K had a mean grain size of 60 μm because of a large grain growth during the HIP process and showed large void swelling. Small grain size is effective in suppressing void swelling due to strong sink effects of grain boundaries for the point defects. The present results indicate that joining at high temperatures may reduce the void swelling resistance of GlidCop copper alloys.

  16. A Model for Surface Induced Growth of Inert Gas Bubbles in Irradiated Copper-Boron Alloys

    SciTech Connect

    Tiwari, G.P.; Ramadasan, E.

    2006-07-01

    A matrix containing inert gas bubbles dilates in direct proportion to the growth experienced by the gas bubbles. This phenomenon is termed as swelling. A model for the swelling induced by the growth of the helium gas bubbles in irradiated copper-boron alloys is presented. The bubbles grow by acquiring vacancies from the external surface, which acts as a source of vacancies. The vacancies reach the surface of the bubbles mainly via lattice diffusion and to a limited extent via diffusion through short-circuiting paths such as grain boundaries and dislocation pipes. The model predicts that overall swelling of the matrix varies as 1.5 power of time. Another consequence of the present model is that the growth rate of a gas bubble varies inversely as the cube of its distance from the external surface. The model has been applied to the data on irradiated copper-boron alloys and found to be in accord with the experimental results. The model is general and can be applied to the growth of all kinds of stationary inert gas bubbles trapped within a crystalline matrix. (authors)

  17. Irradiation effects in rapidly and conventionally solidified alloys. Phase stability in rapidly solidified N i-Nb under Ni ion irradiation

    NASA Technical Reports Server (NTRS)

    1982-01-01

    Two alloy compositions in the Ni-Nb system (Ni60Nb40 and Ni85Nb15) were produced by rapidly quenching from the melt with the piston anvil technique. The Ni60Nb40 was transformed to a metastable, partially crystalline state by heat treatment in a differential scanning calorimeter. The Ni85Nb15 was fully crystalline, with the majority of the grains composed of collections of primary dendrite arms. Both compositions were irradiated with 4 MeV Ni++ ions. The irradiation induced microstructures were examined by transmission electron microscopy and compared with thermally aged samples. The thermal evolution was arrested by ion irradiation in the temperature range studied, by inhibiting the nucleation of the NiNb phase. No irradiation induced voids were observed. It is found that the ion irradiation drives the microstructure along a different path than thermal evolution.

  18. The temperature dependent role of phosphorus and titanium in microstructural evolution of FeCrNi alloys irradiated in FFTF

    NASA Astrophysics Data System (ADS)

    Watanabe, H.; Muroga, T.; Yoshida, N.

    1996-04-01

    The influence of combined addition of phosphorus and titanium on the microstructure of model FeCrNi austenitic alloys irradiated at 700-873 K to 60 dpa has been investigated in comparison with that of a complex austenitic alloy (JPCA-2). At all temperatures, void swelling of the model alloys was suppressed with increasing phosphorus content or co-addition of phosphorus and titanium. The microstructures observed and calculation of defect processes showed that the suppressed void swelling was due to an interaction of phosphorus in solution with defects at lower temperature and needle-like phosphide acting as defect sinks at higher temperature.

  19. Characterization of ion-irradiated ODS Fe-Cr alloys by doppler broadening spectroscopy using a positron beam

    NASA Astrophysics Data System (ADS)

    Parente, P.; Leguey, T.; de Castro, V.; Gigl, T.; Reiner, M.; Hugenschmidt, C.; Pareja, R.

    2015-09-01

    The damage profile of oxide dispersion strengthened steels after single-, or simultaneous triple-ion irradiation at different conditions has been characterized using a low energy positron beam in order to provide information on microstructural changes induced by irradiation. Doppler broadening and coincident Doppler broadening measurements of the positron annihilation line have been performed on different Fe-Cr-(W,Ti) alloys reinforced with Y2O3, to identify the nature and stability of irradiation-induced open-volume defects and their possible association with the oxide nanoparticles. It was found that irradiation induced vacancy clusters are associated with Cr atoms. The results are of highest interest for modeling the damage induced by 14 MeV neutrons in reduced activation Fe-Cr alloys relevant for fusion devices.

  20. Effect of electron- and neutron-irradiation on Fe-Cu model alloys studied by positron annihilation spectroscopy

    NASA Astrophysics Data System (ADS)

    Nagai, Y.; Takadate, K.; Tang, Z.; Ohkubo, H.; Sunaga, H.; Takizawa, H.; Hasegawa, M.

    2011-01-01

    Electron- and neutron-irradiation effects on dilute Fe-Cu model alloys of nuclear reactor pressure vessel steels are studied by positron annihilation spectroscopy. We have found that, not only by high-dose neutron-irradiation but also by low-dose electron-irradiation, the aggregation of Cu atoms and vacancies takes place and the ultrafine Cu precipitates are formed after post-irradiation annealing at 400°C. In spite of large difference in the irradiation doses between the electron- and the neutron-irradiated samples, no significant difference is observed in the isochronal annealing behaviour above 400°C of positron annihilation and micro-hardness, indicating that small amount of extra vacancies enhance the aggregation of Cu atoms in Fe during the annealing-out process of the vacancies.

  1. 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.

  2. 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.

  3. Radiation-induced strengthening in EB welds of Mo-Re alloys during high temperature neutron irradiation

    NASA Astrophysics Data System (ADS)

    Morito, F.; Chakin, V. P.; Danylenko, M. I.; Krajnikov, A. V.

    2011-10-01

    Mo-Re alloys have been known as excellent construction materials with good thermal stability and resistivity for chemical corrosion. These alloys may be fabricated into equipments for various chemical plants and new energy facilities such as fusion reactor. Accordingly it is interesting to elucidate the weldability and radiation performance of Mo-Re alloys in the actual constructions. In this study Mo-Re welds with 16-50% Re exhibited a large radiation-induced strengthening and embrittlement by irradiation at ˜1073 K to ˜5 × 10 21 cm -2 ( E > 0.1 MeV). High temperature neutron irradiation leads to intensive homogeneous nucleation of Re-rich σ-phases in all studied Mo-Re alloys that equalizes the difference in mechanical properties between melting zone, heat-affected zone and base metal. As a result, all parts of as-irradiated welds displayed approximately same level of strength. Therefore, the application of EB welding in Mo-Re constructions operating under high temperature neutron irradiation does not limit lifetime of such constructions.

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

    SciTech Connect

    Gazda, J.

    1998-04-01

    Recent interest in the application of a V-4Cr-4Ti alloy in the ITER prompted an investigation of the effects of low-to-moderate temperature irradiation (<420 C) on the alloy's mechanical properties. Two sets of experiments were conducted. The effects of fast neutron irradiation to {approx_equal}4 dpa at 390 C were investigated in the X530 experiment in the EBR-II reactor. Irradiation with single (4.5-MeV Ni{sup 2}) and dual ion beams (350-keV He{sup +} simultaneously with 4.5-MeV Ni{sup 2}) complemented this study. TEM observations showed the formation of a high density of point-defect clusters and dislocation loops (<30 nm diameter) distributed uniformly in both types of specimens. Mechanical property testing of neutron irradiated material showed embrittlement of the alloy. The deformed microstructure were examined by TEM to determine the causes of embrittlement and revealed dislocation channels propagating through the undeformed matrix. The channels are the sole slip paths and they cause early onset of necking and loss of work-hardening. Based on a review of the available literature, suggestions are made for further research of slip localization in V-4Cr-4Ti alloys.

  5. Effect of Low Temperature Irradiation in ATR On The Mechanical Properties of Ternary V-Cr-Ti Alloys

    SciTech Connect

    Hamilton, Margaret L.; Toloczko, Mychailo B.; Oliver, Brian M.; Garner, Francis A.

    2000-09-01

    Tensile tests and shear punch teats were performed on a variety of vanadium alloys that were irradiated in the Advanced Test Reactor (ATR) at temperatures between 200 and 300 degrees C to doses between 3 and 5 dpa. Tests were performed at room temperature and the irradiation temperature. The results of both the tensile tests and the shear punch tests show that following low temperature irradiation, the yield strength increased by a factor of 3-4 while the ultimate strength increased by a factor of approximately 3.

  6. 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.

  7. 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.

  8. 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

  9. 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

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

    DOE PAGESBeta

    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

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

    DOE PAGESBeta

    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.

  12. 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.

  13. Swelling suppression in phosphorous-modified Fe-Cr-Ni alloys during neutron irradiation

    SciTech Connect

    Lee, E.H.; Packan, N.H.

    1988-01-01

    Phosphorous-containing austenitic alloys in the solution annealed condition were irradiated at 745--760/degree/K. The alloys were variations on Fe--13Cr--15Ni--0.05P with respective additions of 0.8 Si, 0.2 Ti, or 0.8 Si /plus/ 0.2 Ti; also included were low (0.01) and zero P compositions (all values in wt. %). The reference ternary and the two phosphorous-only variations contained little precipitation and numerous voids and swelled rapidly, while the three variants containing P with Si and/or Ti showed little or no void formation and profuse phosphide precipitation. Results indicate that phosphorous in solution alone does not have a major influence on void swelling, whereas fine-scale phosphide precipitation is quite effective at eliminating void formation. The principal mechanism restricting swelling is the effect of the dense precipitate microstructure. These precipitates foster profuse cavity nucleation which in turn dilutes the helium atoms (and more time) in order for individual cavities to surpass their critical size and number of gas atoms necessary for subsequent growth as voids. This mechanism for swelling suppression was not found to be particularly sensitive to moderate variations in either the dislocation or cavity densities; the mechanism is strongest at elevated temperature where the critical quantities are large and is less effective at lower temperatures where the critical quantities are small. 19 refs., 10 figs., 3 tabs.

  14. 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.

  15. 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.

  16. 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.

  17. Construction of radiation-modified phase diagrams under cascade-producing irradiation: application to Zr-Nb alloy

    NASA Astrophysics Data System (ADS)

    Turkin, Anatole A.; Buts, Alexander V.; Bakai, Alexander S.

    2002-10-01

    A general theoretical formalism developed for the description of phase stability alteration in substitutional binary alloys under irradiation is applied to Zr-Nb alloys. We examine the stability of β-Nb precipitates in Zr-Nb alloy subjected to the cascade-producing irradiation. The results of phase stability studies are presented in the form of radiation-modified phase diagram. Evolution of large precipitates (as compared to the size of cascade region) differs from that of small precipitates. In the radiation-modified phase diagram there exists a low temperature boundary for stability of large precipitates, the location of which depends on interface type and displacement rate. Above this boundary large precipitates coarsen with radiation-enhanced rate. Below it the alloy is maintained in a quasi-steady-state of supersaturated solid solution with a population of fine-grained precipitates. The competition between processes of cascade destruction; nucleation and growth of coherent precipitates; and coherency loss can lead to the formation of the distribution of fine-grained precipitates with slowly varying parameters. In particular, such a distribution may form in Zr-Nb alloys under thermal reactor conditions.

  18. Microstructural Characteristics of DU-xMo Alloys with x = 7-12wt%

    SciTech Connect

    Dr. Douglas E. Burkes; Thomas Hartmann; Ramprashad Prabhakaran; Jan-Fong Jue

    2009-06-01

    Microstructural, phase, and impurity analyses of six depleted uranium-molybdenum alloys were obtained using optical metallography, x-ray diffraction, and carbon/nitrogen/oxygen determination. Uranium-molybdenum alloy foils are currently under investigation for the possible conversion of high-power research reactors to accommodate the use of low-enriched uranium fuel. Understanding basic microstructural behavior of these foils is an important consideration in determining the impact of fabrication processes and in anticipating performance of the foils in a reactor. Average grain diameter decreased with increasing molybdenum content, while the degree of mechanical deformation banding increased with increasing molybdenum content. A lattice parameter, which provided information on stress-corrosion cracking behavior, decreased with increasing molybdenum content, and no significant degree of crystallographic ordering was caused by processing and post-processing conditions employed in this study. Impurity concentration, specifically carbon when it is in solution with the host matrix, inhibited the degree of microstructural recrystallization but did not appear to impact other microstructural traits, such as ?-phase retention or lattice parameter.

  19. 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.

  20. Enhancement in anomalous Hall resistivity of Co/Pd multilayer and CoPd alloy by Ga+ ion irradiation

    NASA Astrophysics Data System (ADS)

    Guo, Z. B.; Mi, W. B.; Li, J. Q.; Cheng, Y. C.; Zhang, X. X.

    2014-02-01

    In this paper, we report the effect of Ga+ ion irradiation on anomalous Hall effect (AHE) and longitudinal resistivity (\\rho_{\\textit{xx}}) in [Co(3 Å)/Pd(5 Å)]80 multilayer and Co42Pd58 alloy. 4- and 2-fold increases in anomalous Hall resistivity (\\rho_{\\textit{AH}}) in the Co/Pd multilayer and CoPd alloy have been observed after irradiations at doses of 2.4\\times 10^{15} and 3.3\\times 10^{15}\\ \\text{ions/cm}^{2} , respectively. Skew scattering and side jump contributions to AHE have been analyzed based on the scaling relationship \\rho_{\\textit{AH}}=a\\rho_{\\textit{xx}}+b\\rho_{\\textit{xx}}^{2} . For the Co/Pd multilayer, AHE is mainly affected by ion irradiation-induced interface diffusion and defects. For the CoPd alloy, the increase in doses above 1.5\\times 10^{15}\\ \\text{ions/cm}^{2} induces a sign change in skew scattering, followed by the skew scattering contribution to AHE overwhelming the side jump contribution, this phenomenon should be attributed to irradiation-induced defects and modifications in chemical ordering.

  1. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    SciTech Connect

    D D. Keiser, Jr.; A. B. Robinson; D. E. Janney; J. F. Jue

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels.

  2. Radiation effects on microstructure and hardness of a titanium aluminide alloy irradiated by helium ions at room and elevated temperatures

    NASA Astrophysics Data System (ADS)

    Wei, Tao; Zhu, Hanliang; Ionescu, Mihail; Dayal, Pranesh; Davis, Joel; Carr, David; Harrison, Robert; Edwards, Lyndon

    2015-04-01

    A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5 MeV helium ions to a fluence of 5 × 1021 ion m-2 (5000 appm) with a dose of about 1 dpa (displacements per atom). A uniform helium ion stopping damage region about 17 μm deep from the target surface was achieved by applying an energy degrading wheel. Radiation damage defects including helium-vacancy clusters and small helium bubbles were found in the microstructure of the samples irradiated at room temperature. With increasing irradiation temperature to 300 °C and 500 °C helium bubbles were clearly observed in both the α2 and γ phases of the irradiated microstructure. By means of nanoindentation significant irradiation hardening was measured. For the samples irradiated at room temperature the hardness increased from 5.6 GPa to 8.5 GPa and the irradiation-hardening effect reduced to approximately 8.0 GPa for the samples irradiated at 300 °C and 500 °C.

  3. Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

    NASA Astrophysics Data System (ADS)

    Nishi, Takakuni; Hashimoto, N.; Ohnuki, S.; Yamamoto, T.; Odette, G. R.

    2011-10-01

    Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn-Ni-Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn-Ni-Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

  4. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    SciTech Connect

    D. D. Keiser, Jr.; A. B. Robinson; D. E. Janney; J. F. Jue

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with either Al-0.2Si or 4043 Al (~4.8% Si) alloy matrix in the as-fabricated and/or as-irradiated condition using optical metallography and/or scanning electron microscopy. Fuel plates with either matrix can have Si-rich layers around the U-7Mo particles after fabrication, and during irradiation these layers were observed to grow in thickness and to become Si-deficient in some areas of the fuel plates. For the fuel plates with 4043 Al, this was observed in fuel plate areas that were exposed to very aggressive irradiation conditions.

  5. 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.

  6. 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.

  7. Nonequilibrium segregation and phase instability in alloy films during elevated-temperature irradiation in a high-voltage electron microscope

    SciTech Connect

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

    1984-05-01

    The effects of defect-production rate gradients, caused by the radial nonuniformity in the electron flux distribution, on solute segregation and phase stability in alloy films undergoing high-voltage electron-microscope (HVEM) irradiation at high temperatures are assessed. Two-dimensional (axially symmetric) compositional redistributions were calculated, taking into account both axial and transverse radial defect fluxes. It was found that when highly focused beams were employed radiation-induced segregation consisted of two stages: dominant axial segregation at the film surfaces at short irradiation times and competitive radial segregation at longer times. The average alloy composition within the irradiated region could differ greatly from that irradiated with a uniform beam, because of the additional atom transport from or to the region surrounding the irradiated zone under the influence of radial fluxes. As a result, damage-rate gradient effects must be taken into account when interpreting in-situ HVEM observations of segregation-induced phase instabilities. The theoretical predictions are compared with experimental observations of the temporal and spatial dependence of segregation-induced precipitation in thin films of Ni-Al, Ni-Ge and Ni-Si solid solutions.

  8. 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.

  9. Nanopatterns induced by pulsed laser irradiation on the surface of an Fe-Al alloy and their magnetic properties

    SciTech Connect

    Yoshida, Yutaka; Oosawa, Kazuya; Watanabe, Seiichi; Kaiju, Hideo; Kondo, Kenji; Ishibashi, Akira; Yoshimi, Kyosuke

    2013-05-06

    We have studied nanopatterns induced by nanosecond pulsed laser irradiation on (111) plane surfaces of a polycrystalline iron-aluminum alloy and evaluated their magnetic properties. Multiple nanosecond pulsed laser irradiation induces a wavelength-dependent surface transformation of the lattice structure from a B2-type to a supersaturated body centered cubic lattice. The selective formation of surface nanopatterns consisting of holes, stripes, polygonal networks, and dot-like nanoprotrusions can be observed. Furthermore, focused magneto-optical Kerr effect measurements reveal that the magnetic properties of the resultant nanostructured region changes from a paramagnetic to a ferromagnetic phase in accordance with the number of laser pulses.

  10. Electron irradiation-induced defects in Mo-diluted FeCrNi austenitic alloy during void swelling incubation

    NASA Astrophysics Data System (ADS)

    Wang, B. Y.; Lu, E. Y.; Zhang, C. X.; Xu, Q.; Jin, S. X.; Zhang, P.; Cao, X. Z.

    2016-01-01

    The microstructural features and the effect of Mo addition in FeCrNi austenitic alloy during incubation period were investigated using positron annihilation technique and micro- Vickers Hardness. The electron irradiation, which could induce vacancy defects in material, was performed at room temperature up to the dose of 1.7×10-4 and 5×10-4 dpa, respectively. The defect concentration was estimated about 10-4-10-7 though the standard trapping model. The added Mo atoms could trap vacancies to form Mo-vacancy complexes, which may restrain the migration and growth of vacancy defects during electron irradiation. In addition, the microstructural evolution during electron radiation resulted in hardening, while the added Mo might improve the hardening property of the alloy.

  11. Effect of laser irradiation conditions on the laser welding strength of cobalt-chromium and gold alloys.

    PubMed

    Kikuchi, Hisaji; Kurotani, Tomoko; Kaketani, Masahiro; Hiraguchi, Hisako; Hirose, Hideharu; Yoneyama, Takayuki

    2011-09-01

    Using tensile tests, this study investigated differences in the welding strength of casts of cobalt-chromium and gold alloys resulting from changes in the voltage and pulse duration in order to clarify the optimum conditions of laser irradiation for achieving favorable welding strength. Laser irradiation was performed at voltages of 150 V and 170 V with pulse durations of 4, 8, and 12 ms. For cobalt-chromium and gold alloys, it was found that a good welding strength could be achieved using a voltage of 170 V, a pulse duration of 8 ms, and a spot diameter of 0.5 mm. However, when the power density was set higher than this, defects tended to occur, suggesting the need for care when establishing welding conditions. PMID:21959656

  12. 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.

  13. 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.

  14. 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

  15. Microstructure of RERTR Du-alloys irradiated with krypton ions up to 100 dpa.

    SciTech Connect

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

    2011-04-01

    The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al){sub 3}, (U, Mo)(Si, Al){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43}, and UAl{sub 4}. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 C to doses up to {approx}100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al){sub 3} and UAl{sub 4} phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al){sub 3} and UMo{sub 2}Al{sub 20} phases become amorphous at 1 and {approx}2 dpa, respectively, and show no evidence of voids at 100 dpa. The U{sub 6}Mo{sub 4}Al{sub 43} phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. Direct magnetic patterning due to the generation of ferromagnetism by selective ion irradiation of paramagnetic FeAl alloys.

    PubMed

    Menéndez, Enric; Liedke, Maciej Oskar; Fassbender, Jürgen; Gemming, Thomas; Weber, Anja; Heyderman, Laura J; Rao, K V; Deevi, Seetharama C; Suriñach, Santiago; Baró, Maria Dolors; Sort, Jordi; Nogués, Josep

    2009-02-01

    Sub-100-nm magnetic dots embedded in a non-magnetic matrix are controllably generated by selective ion irradiation of paramagnetic Fe(60)Al(40) (atomic %) alloys, taking advantage of the disorder-induced magnetism in this material. The process is demonstrated by sequential focused ion beam irradiation and by in-parallel broad-beam ion irradiation through lithographed masks. Due to the low fluences used, this method results in practically no alteration of the surface roughness. The dots exhibit a range of magnetic properties depending on the size and shape of the structures, with the smallest dots (<100 nm) having square hysteresis loops with coercivities in excess of micro(0)H(C) = 50 mT. Importantly, the patterning can be fully removed by annealing. The combination of properties induced by the direct magnetic patterning is appealing for a wide range of applications, such as patterned media, magnetic separators, or sensors. PMID:19089839

  2. Phase transformation studies in unirradiated and proton beam irradiated Ni-Ti alloy between 25 and 100°C

    NASA Astrophysics Data System (ADS)

    Ayub, Rana; Afzal, Naveed; Ahmad, R.

    2012-06-01

    The stress-induced phase transformation characteristics of unirradiated and proton beam irradiated NiTi alloy were investigated at different tests temperatures. The wire-shaped NiTi specimens were irradiated by 2 MeV proton beam for 30 min at room temperature to a flux of 1019 protons/m2 s. Engineering stress-strain (S-S) curves of both unirradiated and irradiated specimens were obtained using a materials testing machine at 25, 50, 75 and 100°C. The results indicate a single-stage phase transformation from austenite to martensite (B2-B19‧) in unirraidated specimens at all the test temperatures. In contrast, in the case of the irradiated specimens, a two-stage austenite-rhombohedral-martensite (B2-R-B19‧) phase transformation is observed at 25 and 50°C. The B2-R-B19‧ phase transformation, however, is found to change into B2-B19‧ transformation at 75 and 100°C. The stress required to initiate the B19‧ phase transformation (σMS) and the plateau range are found to be lower in irradiated specimens compared with those of the unirradiated specimens. The results obtained are discussed on the basis of the formation of Ni4Ti3 precipitates in irradiated specimens and their consequences on the phase transformations.

  3. 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.

  4. Post-irradiation annealing behavior of neutron-irradiated FeCu, FeMnNi and FeMnNiCu model alloys investigated by means of small-angle neutron scattering

    NASA Astrophysics Data System (ADS)

    Bergner, F.; Ulbricht, A.; Lindner, P.; Keiderling, U.; Malerba, L.

    2014-11-01

    Neutron irradiation of reactor pressure vessel steels gives rise to the formation of thermodynamically stable and unstable nano-features. The present work is focused on the stability of Cu-, Mn- and Ni-containing solute clusters in model alloys exposed to post-irradiation annealing. Fe0.1Cu, Fe1.2Mn0.7Ni and Fe1.2Mn0.7Ni0.1Cu (wt%) model alloys irradiated up to neutron exposures of 0.1 and 0.19 dpa (displacements per atom) were annealed at stepwise increasing temperatures in the range from 300 °C (i.e. near irradiation temperature) to 500 °C and characterized by means of small-angle neutron scattering (SANS). We have found characteristic differences in the annealing behavior of the alloys. In particular, there is a non-trivial (synergistic-antagonistic) interplay of Mn/Ni and Cu.

  5. 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.

  6. Effects of dpa rate on swelling in neutron-irradiated Fe-Cr and Fe-Cr-Mo alloys

    NASA Astrophysics Data System (ADS)

    Okita, T.; Sekimura, N.; Garner, F. A.

    2011-10-01

    Data are presented on the void swelling of three model Fe-Cr ferritic alloys following irradiation in TEM packets in FFTF-MOTA over the range 373-600 °C and a wide range of dpa rates. It is shown that raising the chromium level decreases the steady-state swelling rate at ˜420 °C. Addition of Mo to the Fe-12Cr alloy does not change the swelling rate significantly but does lead to an apparent swelling of ˜3% that arises from the radiation-accelerated formation of Chi phase. Swelling tends to decrease with increasing irradiation temperature for all three alloys. It is shown that the sensitivity of swelling to dpa rate expresses itself not at the various packet positions in FFTF, each with their characteristic nominal dpa rates, but also in response to variations in dpa rate along the length of the packet containing the specimens. The latter introduces second-order uncertainties in determination of the dpa levels and dpa rates, but these are not sufficient to obscure the major conclusion concerning dpa rate and composition.

  7. 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.

  8. 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

  9. 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

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

    DOE PAGESBeta

    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

  11. 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.

  12. Silicon`s role in determining swelling in neutron-irradiated Fe-Cr-Ni-Si alloys

    SciTech Connect

    Sekimura, N.; Garner, F. A.; Newkirk, J.W.

    1991-11-01

    Two silicon-modified alloy series, one based on Fe-15Cr-20Ni and another based on Fe-15Cr-25Ni were irradiated at target temperatures between 399 and 649{degree}C in EBR-II. The influence of silicon on swelling is more complex than previously envisioned and indicates that silicon plays two or more competing roles while in solution. Radiation-induced formation of {gamma}{prime} (Ni{sub 3}Si) precipitates is dependent on silicon and nickel content, as well as temperature. Precipitation of {gamma}{prime} appears to play only a minor role in void formation.

  13. Dislocation loop nucleation and growth in zirconium-2.5 wt% niobium alloy during 1 MeV electron irradiation

    NASA Astrophysics Data System (ADS)

    Buckley, S. N.; Manthorpe, S. A.

    1980-05-01

    A high voltage electron microscope has been used to study loop formation in thin foils of Zr, Zircaloy and two Zr-Nb alloys. The effects of varying irradiation temperature and displacement dose were examined in the light of nucleation and growth theories. It was found that Nb solute enhanced loop densities by up to more than an order of magnitude but loop growth at 773 K was more than proportionately retarded so that total point defect precipitation is diminished. The temperature dependence of loop density corresponds to an activation energy 0.36 eV with <0.1 eV solute binding.

  14. 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.

  15. Characterization of neutron-irradiated ferritic model alloys and a RPV steel from combined APT, SANS, TEM and PAS analyses

    NASA Astrophysics Data System (ADS)

    Meslin, E.; Lambrecht, M.; Hernández-Mayoral, M.; Bergner, F.; Malerba, L.; Pareige, P.; Radiguet, B.; Barbu, A.; Gómez-Briceño, D.; Ulbricht, A.; Almazouzi, A.

    2010-11-01

    Understanding the behavior of reactor pressure vessel (RPV) steels under irradiation is a mandatory task that has to be elucidated in order to be able to operate safely a nuclear power plant or to extend its lifetime. To build up predictive tools, a substantial experimental data base is needed at the nanometre scale to extract quantitative information on neutron-irradiated materials and to validate the theoretical models. To reach this experimental goal, ferritic model alloys and French RPV steel were neutron irradiated in a test reactor at an irradiation flux of 9 × 10 17 nm -2 s, doses from 0.18 to 1.3 × 10 24 nm -2 and 300 °C. The main goal of this paper is to report the characterization of the radiation-induced microstructural change in the materials by using the state-of-the-art of characterization techniques available in Europe at the nanometre scale. Possibilities, limitations and complementarities of the techniques to each other are highlighted.

  16. X-ray diffraction analysis of secondary phases in zirconium alloys before and after neutron irradiation at the MARS synchrotron radiation beamline

    NASA Astrophysics Data System (ADS)

    Béchade, J.-L.; Menut, D.; Doriot, S.; Schlutig, S.; Sitaud, B.

    2013-06-01

    To further advance understanding of microstructural evolution in zirconium alloys for high burnup applications in PWR, it is important to obtain precise characterizations of second-phase particles present in the bulk alloys as a function of the neutron-irradiation fluence. X-ray diffraction from a synchrotron radiation source (SOLEIL) was used to identify and follow the evolution of second-phase particles that are in very small volume fractions in two zirconium alloys: Zy-4 in stress-relieved metallurgical state before irradiation and Zr-1%Nb (M5®) in recrystallized metallurgical state before and after neutron irradiation. Despite the fact the neutron irradiated sample is not in the as-irradiated state due to a thermal treatment and creep test performed after irradiation, interesting results have been obtained on secondary phases using XRD techniques. Analyses have been performed at the MARS beamline, fully dedicated to advanced structural and chemical characterizations of radioactive matter. A first proof of the improvement brought by these new analyses performed at the MARS beamline is given with Zr(Fe, Cr)2 precipitates found in unirradiated Zy-4 alloy in stress relieved metallurgical state, highly textured and displaying significant residual stresses and numerous dislocations: lattice parameters, crystallite size and microstrains (line profile analysis using the Williamson-Hall method after correction for instrumental broadening) have been estimated with very good accuracy. Then, second phase particles of the Zr-1%Nb alloy (M5®) have been analyzed before and after irradiation. For the Zr(Fe, Nb)2 Laves phase, the diffraction line disappeared after neutron irradiation. For β-Nb phases, the evolution of diffraction peaks clearly show the convolution of two phenomena: in one hand slight decreases in Nb content for native β-Nb particles and on the other hand irradiation-enhanced precipitation of nano-sized needle-like β-Nb particles. To our knowledge, this is

  17. 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.

  18. 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.

  19. The Effect of H and He on Irradiation Performance of Fe and Ferritic Alloys

    SciTech Connect

    James F. Stubbins

    2010-01-22

    This research program was designed to look at basic radiation damage and effects and mechanical properties in Fe and ferritic alloys. The program scope included a number of materials ranging from pure single crystal Fe to more complex Fe-Cr-C alloys. The range of materials was designed to examine materials response and performance on ideal/model systems and gradually move to more complex systems. The experimental program was coordinated with a modeling effort. The use of pure and model alloys also facilitated the ability to develop and employ atomistic-scale modeling techniques to understand the inherent physics underlying materials performance

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Ion mass dependence of irradiation-induced local creation of ferromagnetism in Fe60Al40 alloys

    NASA Astrophysics Data System (ADS)

    Fassbender, J.; Liedke, M. O.; Strache, T.; Möller, W.; Menéndez, E.; Sort, J.; Rao, K. V.; Deevi, S. C.; Nogués, J.

    2008-05-01

    Ion irradiation of Fe60Al40 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+ , Kr+ , and Xe+ ), 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+ and Ne+ ), 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.

  6. 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.

  7. 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.

  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

    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.

  9. 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.

  10. 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.

  11. 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.

  12. Synthesis of carbon-supported PtRh random alloy nanoparticles using electron beam irradiation reduction method

    NASA Astrophysics Data System (ADS)

    Matsuura, Yoshiyuki; Seino, Satoshi; Okazaki, Tomohisa; Akita, Tomoki; Nakagawa, Takashi; Yamamoto, Takao A.

    2016-05-01

    Bimetallic nanoparticle catalysts of PtRh supported on carbon were synthesized using an electron beam irradiation reduction method. The PtRh nanoparticle catalysts were composed of particles 2-3 nm in size, which were well dispersed on the surface of the carbon support nanoparticles. Analyses of X-ray diffraction and scanning transmission electron microscopy-energy-dispersive X-ray spectroscopy revealed that the PtRh nanoparticles have a randomly alloyed structure. The lattice constant of the PtRh nanoparticles showed good correlation with Vegard's law. These results are explained by the radiochemical formation process of the PtRh nanoparticles. Catalytic activities of PtRh/C nanoparticles for ethanol oxidation reaction were found to be higher than those obtained with Pt/C.

  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. 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.

  16. 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.

  17. 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.

  18. Mechanistic understanding of irradiation-induced corrosion of zirconium alloys in nuclear power plants: Stimuli, status, and outlook

    SciTech Connect

    Johnson, A.B. Jr.; Ishigure, K.; Nechaev, A.F.; Reznichenko, E.A.; Cox, B.; Lemaignan, C.; Petrik, N.G.

    1990-05-01

    Failures in the basic materials used in nuclear power plants continue to be costly and insidious, despite increasing industry vigilance to catch failures before they degrade safety. For instance, the overall costs to the US industry from materials problems could amount to as much as $10 billion annually. Moreover, estimates indicate that the cost of a pipe failure in a nuclear plant is one hundred times greater than the cost of a similar failure in a coal-fired plant. There are important practical stimuli and much scope for further understanding of the effects of irradiation on Zr-alloys (and other materials used in nuclear installations) by careful experimentation. Moreover, these studies need to address the effect of irradiation on all components of heterogeneous systems: the metal, the oxide and the environment, and especially those processes recurring at the interphases between these components. The present paper is aimed at providing specialists with some systematic information on the subject and with important considerations on the key items for further experimentation.

  19. Enhancement of Curie Temperature (T c) and Magnetization of Fe-Ni Invar alloy Through Cu Substitution and with He+2 Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Khan, Sajjad Ahmad; Ziya, Amer Bashir; Ibrahim, Ather; Atiq, Shabbar; Usman, Muhammad; Ahmad, Naseeb; Shakeel, Muhammad

    2016-04-01

    The magnetic properties of ternary Fe-Ni-Cu invar alloys are affected by ion irradiation, which goes on increasing with increasing ion fluence (Φ), and by increasing Cu content. In the present study, the ions used are He+2 with 2 MeV energy and with 1 × 1013 cm-2, 1 × 1014 cm-2, 5 × 1014 cm-2, 1 × 1015 cm-2 and 5 × 1015 cm-2 fluence (dose) for irradiation purpose. The face centered cubic structure of the alloy was investigated after ion irradiation using x-ray diffraction (XRD) and found unchanged. However, the peaks become broader with increasing ion dose. Additionally, the lattice fluctuations were observed in XRD study. Curie temperature (T c) is also increased after irradiation. Many factors are considered here for the reason for increasing T c, such as the stopping of incident ions, atomic mixing effect at micro scale level owing to ion irradiation, which might change local concentration and ordering already reported in diffuse scattering, and as a result the Fe-Fe interatomic distance and the Fe-Fe coupling are changed. A comparative study shows that the effect of irradiation on T c and magnetization with increasing ion fluence is more distinctive than the addition of Cu.

  20. Time dependency of the hydrophilicity and hydrophobicity of metallic alloys subjected to femtosecond laser irradiations

    NASA Astrophysics Data System (ADS)

    Bizi-bandoki, P.; Valette, S.; Audouard, E.; Benayoun, S.

    2013-05-01

    Surfaces of metallic alloys were laser-processed with femtosecond laser pulses of 800 nm, with different power densities. The effect of time on the wettability of these surfaces was investigated. A multi-scale roughness made of undulations was created after the laser processing. This specific surface topography allowed the occurrence of a Wenzel's state. This state clearly explains the high hydrophilicity and hydrophobicity observed respectively one day after laser treatment and several days later. The change from hydrophilicity to hydrophobicity occurred over time and is due to surface chemistry modifications. The creation of new hydrophobic functional groups on aluminum alloy surface, for example, was proposed to be responsible for the hydrophobic behavior observed on these surfaces.

  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. Effects of thermal annealing and ion irradiation on the properties and microstructures of copper alloys. [AMZIRC(Cu-Zr) and AMAX-MZC(Cu-Cr-Zr-Mg)

    SciTech Connect

    Zinkle, S.J.

    1985-01-01

    High strength, high conductivity copper alloys have recently been developed for a variety of technologically challenging applications, including proposed nuclear fusion reactors. The purpose of this research is to examine the effects of high temperatures and radiation levels on the microstructural stability of two commercial high-strength, high conductivity copper alloys, AMZIRC (Cu-Zr) and AMAX-MZC (Cu-Cr-Zr-Mg). The recrystallization temperature of both AMZIRC and AMAX-MZC was measured to be 475/sup 0/C for a 1 hour anneal, and was estimated to be 320/sup 0/C for a 20 year anneal which corresponds to a recrystallization energy of 3.0 eV. The yield strength of AMZIRC and AMAX-MZC decreased by at least a factor of two upon recrystallization. No significant void formation was observed in pure copper or in the AMZIRC and MZC copper alloys following ion irradiation at temperatures of 100-500/sup 0/C (0.28-0.59 T/sub M/) to peak damage levels of 40 dpa. Instead, stacking fault tetrahedra (SFT) were detected. The absence of voids following ion irradiation is attributed to the low oxygen content of the copper and copper alloys (<5 ppm). Ion irradiation of cold-worked plus aged AMZIRC and AMAX-MZC caused an effective shift in the recrystallization temperature to a lower value by 75 to 150/sup 0/C compared with thermal anneal results. A simple model based on radiation-enhanced diffusion predicts a radiation-enhanced recrystallization temperature of 150/sup 0/C for these alloys for a 2 year neutron irradiation.

  3. 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.

  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 study on neutron irradiated pure iron and RPV model alloys

    NASA Astrophysics Data System (ADS)

    Hernández-Mayoral, M.; Gómez-Briceño, D.

    2010-04-01

    The radiation induced microstructure was examined by Transmission Electron Microscopy in Fe, FeCu, FeMnCuNi, FeMnNi and a Reactor Pressure Vessel steel that were neutron irradiated to 0.026, 0.051, 0.10 and 0.19 dpa at 300 °C. The effect of dose and composition on defect accumulation and microstructure evolution was investigated. The damaged microstructure consisted in the presence of dislocation loops of interstitial type. The presence of voids was also studied in pure iron. Results on density, size and Burgers vector of radiation induced dislocation loops showed that the evolution of the interstitial component of the neutron irradiation induced microstructure was strongly affected by the presence of solutes such as Cu, Mn and Ni. Density and size increased with increasing dose in all the materials, while the effect of solutes is clearly to decrease the size of defects compared to pure iron. It has been observed that, for the same irradiation dose, the defect size decreases as the material becomes more complex, with the extreme case of the RPV steel where no defects were observed at any of the irradiation doses studied.

  6. 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.

  7. 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.

  8. 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.

  9. The formation of bubbles in Zr alloys under Kr ion irradiation

    SciTech Connect

    Birtcher, R.C.; Pagano, L. Jr.; Motta, A.T.

    1996-06-01

    The authors report here a study of Kr ion implantation and the resultant bubble formation in Zr and Zr alloys, including Zircaloy-2 and Zircaloy-4. Implantations into thin foils were performed in the HVEM/Tandem facility at Argonne National Laboratory at temperatures between 300 to 800 C and to doses up to 2 x 10{sup 16} ion.cm{sup {minus}2}. Bulk specimens were implanted in an ion-beam chamber and then thinned for viewing by TEM. In thin foils, only small bubbles (30--100 {angstrom}) were formed at all temperatures with the exception of the Cr-rich V alloy where bubbles of 130 {angstrom} bubbles were formed. Bulk samples implanted at 300 C contained a bubble morphology similar to that observed after implantation into thin foils. However, at high temperatures (500--800 C) large faceted bubbles (up to 300 {angstrom}) were produced in bulk specimens. The results indicate that bubble formation and evolution below 500 C is controlled by gas concentration, while it is controlled by bubble mobility at high temperatures.

  10. Roles of vacancy/interstitial diffusion and segregation in the microchemistry at grain boundaries of irradiated Fe-Cr-Ni alloys

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

    This work presents 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. 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, 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.

  11. 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.

  12. 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.

  13. Spectral effects in low-dose fission and fusion neutron irradiated metals and alloys

    SciTech Connect

    Heinisch, H.L.; Atkin, S.D.; Martinez, C.

    1986-04-01

    Flat miniature tensile specimens were irradiated to neutron fluences up to 9 x 10/sup 22/ n/m/sup 2/ in the RTNS-II and in the Omega West Reactor. Specimen temperatures were the same in both environments, with runs being made at both 90/sup 0/C and 290/sup 0/C. The results of tensile tests on AISI 316 stainless steel, A302B pressure vessel steel and pure copper are reported here. The radiation-induced changes in yield strength as a function of neutron dose in each spectrum are compared. The data for 316 stainless steel correlate well on the basis of displacements per atom (dpa), while those for copper and A302B do not. In copper the ratio of fission dpa to 14 MeV neutron dpa for a given yield stress change is about three to one. In A302B pressure vessel steel this ratio is more than three at lower fluences, but the yield stress data for fission and 14 MeV neutron-irradiated A302B steel appears to coalesce or intersect at the higher fluences.

  14. Anisotropic dislocation loop distribution in alloy T91 during irradiation creep

    NASA Astrophysics Data System (ADS)

    Xu, Cheng; Was, Gary S.

    2014-11-01

    Ferritic-martensitic (FM) steel T91 was subjected to irradiation with 3 MeV protons while under load at stresses of 100 MPa, 180 MPa, and 200 MPa. Transmission electron microscopy of crept T91 samples was conducted to determine the nature of the loops and to quantify the edge-on dislocation loop density as a function of their orientation to the tensile axis. Loops were determined to be interstitial in nature of the type ao<1 0 0>. The distribution of ao<1 0 0> loops was found to be strongly anisotropic in which loops preferentially formed with their normal in the tensile direction. An empirical correlation was developed to describe the relationship between the orientation of the loop normal to the tensile axis and the local dislocation loop density. Loop anisotropy was found to increase as a function of the applied stress and their strain contributions relative to total creep strain were calculated at 4-11% of the total creep strain. Results from this study provide a means for quantifying empirically measured anisotropic microstructure and including it in irradiation creep mechanisms for FM steels.

  15. New method for detection of Li inside He bubbles formed in B10-alloyed steel after neutron irradiation.

    PubMed

    Klimenkov, M; Möslang, A; Materna-Morris, E

    2013-03-01

    Electron energy loss spectroscopy (EELS) was used to detect and study the spatial distribution on the nanoscale of He and Li in boron-alloyed steel after neutron irradiation. Li and He are the products of the (10)B(n, α)(7)Li nuclear transmutation reaction and knowledge of their distribution is important to understand their influence on mechanical properties. Here, a new method is presented for the direct detection of Li in Fe, which is based on the analysis of the plasmon structure in EELS spectra. Li drops or particles in He bubbles show pronounced Li plasmon line at 10eV which can be extracted from the Fe/Cr plasmon. The Gaussian or linear interpolation of the Fe/Cr plasmon and its subtraction allows for the calculation of Li and He two-dimensional maps and the study their spatial distribution. The analysis of Li plasmon fine structure allows imaging surface effects in the Li drops. PMID:23332433

  16. Microstructure of TiB{sub 2}/carbon steel surface-alloyed materials fabricated by high-energy electron beam irradiation

    SciTech Connect

    Euh, K. Lee, S.; Shin, K.

    1999-12-01

    The processing and the microstructural analysis of TiB{sub 2}/carbon steel surface-alloyed materials using the irradiation of a high-energy electron beam were investigated in this study. The mixtures of TiB{sub 2} powders and flux were deposited on a plain carbon steel substrate, and then electron beam was irradiated on these mixtures using an electron beam accelerator. The microstructure of the irradiated surface layer was composed of a melted region, an interfacial region, a coarse-grained heat-affected zone (HAZ), and a fine-grained HAZ. A few residual micropores were found in the melted region of the specimen processed without flux because of irregular thermal transfer, but their number was decreased in the specimens processed with a considerable amount of flux. As a result of irradiation, the Ti content was homogeneously maintained throughout the melted region, whose hardness was greatly improved. This was associated with the microstructural modification including the segregation of Ti and B along solidification cell boundaries and the formation of fine Ti(C, N) particles. The proper flux mix ratio was 15 to 30% to obtain excellent surface alloying and a homogeneous microstructure.

  17. 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.

  18. 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.

  19. Damage recovery in electron-irradiated FeNiCr alloys

    NASA Astrophysics Data System (ADS)

    Mekhrabov, Amdulla O. O.; Sato, Eiichi; Shimotomai, Michio; Doyama, Masao; Iwata, Tadao; Kawanishi, Toshio

    1985-08-01

    Two martensitic steels (Fe-8.5 Cr-0.9 MoVNb and Fe-9.0 Cr-1.8 MoVNb), one ferritic steel (Fe-17 Cr-1.5 MoVNb), 316 stainless steel and Ni 3Fe were irradiated with 2 MeV electrons to 1× 10 23 e/m 2. The damage recovery was studied by positron annihilation and differential scanning calorimetry (DSC). Results suggest that vacancies migrate and disappear below 330 K in the martensitic phase. Vacancies in the ferritic and austenitic phases disappear around 420 K, enhancing the precipitation. Ordered Ni 3Fe has three DSC annealing stages between 400 and 600 K.

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Irradiation hardening of Fe-9Cr-based alloys and ODS Eurofer: Effect of helium implantation and iron-ion irradiation at 300 °C including sequence effects

    NASA Astrophysics Data System (ADS)

    Heintze, C.; Bergner, F.; Hernández-Mayoral, M.; Kögler, R.; Müller, G.; Ulbricht, A.

    2016-03-01

    Single-beam, dual-beam and sequential iron- and/or helium-ion irradiations are widely accepted to emulate more application-relevant but hardly accessible irradiation conditions of generation-IV fission and fusion candidate materials for certain purposes such as material pre-selection, identification of basic mechanisms or model calibration. However, systematic investigations of sequence effects capable to critically question individual approaches are largely missing. In the present study, sequence effects of iron-ion irradiations at 300 °C up to 5 dpa and helium implantations up to 100 appm He are investigated by means of post-irradiation nanoindentation of an Fe9%Cr model alloy, ferritic/martensitic 9%Cr steels T91 and Eurofer97 and oxide dispersion strengthened (ODS) Eurofer. Different types of sequence effects, both synergistic and antagonistic, are identified and tentative interpretations are suggested. It is found that different accelerated irradiation approaches have a great impact on the mechanical hardening. This stresses the importance of experimental design in attempts to emulate in-reactor conditions.

  8. Effect of FFTF irradiation on tensile properties of P- and Ti-modified model austenitic alloys with small amounts of boron

    NASA Astrophysics Data System (ADS)

    Kurishita, H.; Muroga, T.; Watanabe, H.; Yoshida, N.; Kayano, H.; Hamilton, M. L.

    1994-09-01

    The interactive effects of P and Ti additions, {helium}/{dpa} ratio, irradiation and test temperature on postirradiation tensile properties of a model Fe-16Cr-17Ni alloy were investigated. Miniature tensile specimens containing 64 and 522 appm 10B (0.75 and 3.8 {He}/{dpa} ratio), with and without 0.1 wt% P and 0.25 wt% Ti additions, were irradiated to 33 dpa at 703, 793 and 874 K in the Fast Flux Test Facility (FFTF). They were deformed at 300, 473 and 673 K. For all the alloy conditions the irradiation at 703 K increases the yield stress and decreases the uniform elongation significantly. An effect of 10B addition occurs, which is not related to helium generation, but it is small compared to the effects of P and Ti additions. Additions of P and Ti, especially their simultaneous addition, cause a significant strengthening but no appreciable change in uniform elongation. The cause of the strengthening and the observed changes in uniform elongation are discussed.

  9. Kinetic of solute clustering in neutron irradiated ferritic model alloys and a French pressure vessel steel investigated by atom probe tomography

    NASA Astrophysics Data System (ADS)

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

    2010-04-01

    The embrittlement of reactor pressure vessel steels under neutron irradiation is partly due to the formation of solute clusters. To gain more insight into their formation mechanisms, ferritic model alloys (low copper Fe-0.08 at.% Cu, Fe-0.09 Cu-1.1 Mn-0.7 Ni (at.%), and a copper free Fe-1.1 Mn-0.7 Ni (at.%)) and a French 16MND5 reactor pressure vessel steel, were irradiated in a test reactor at two fluxes of 0.15 and 9 × 10 17 n( E> 1 MeV) m -2 s -1 and at increasing doses from 0.18 to 1.3 × 10 24 n( E> 1 MeV) m -2. Atom probe tomography analyses revealed that nanometer-size solute clusters were formed during irradiation in all the materials, even in the copper free Fe-1.1 Mn-0.7 Ni (at.%) alloy. It should be noted that solute segregation in a low-Ni ferritic material was never reported before in absence of the highly insoluble copper impurity. The manganese and nickel segregation is suggested to result from a radiation-induced mechanism.

  10. Evolution of cavity microstructure in ion-irradiated 316 SS and Fe-20Ni-15Cr alloy. [Fe-20Ni-15Cr

    SciTech Connect

    Kohyama, A.; Loomis, B.; Ayrault, G.; Igata, N.

    1984-04-01

    The effect of helium implantation schedule on the evolution of the cavity microstructure in 316 SS and Fe-20Ni-15Cr alloy during heavy-ion irradiation was investigated for damage levels up to 100 dpa with three helium injection schedules, i.e., 15 appm He/dpa, 50 appm He/dpa and 15 appm He preinjected prior to Ni/sup +/ ion irradiation. In the case of the dual-ion irradiated specimens, there was a trend towards a saturation of the swelling with increasing damage level that was closely related to the saturation and subsequent decrease of the number density of the small cavities (<8 nm diameter for 316 SS and <4 nm for Fe-20Ni-15Cr) and the dislocation density. The bi-model cavity size distribution that was determined for the Fe-20Ni-15Cr alloy at 40 dpa changed to uni-modal distribution at 65 and 100 dpa. Radiation-induced segregation was high and this was related to the saturation tendency for the swelling.

  11. Development of new reactor fuel materials: hydrogenation properties of UThZr alloys and neutron irradiation effects on their hydrides

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takuya; Suwamo, Hadi; Kayano, Hideo; Yamawaki, Michio

    1997-08-01

    Hydrogen absorption properties of four UThZr alloys with compositions of 2:1:6, 1:1:4, 1:2:6 and 1:4:10 in U:Th:Zr ratio were examined for developing a new UTh mixed hydride fuel. It was revealed by hydrogen absorption measurement that for all the specimens hydrogen capacities normalized by the sum of Th and Zr contents were similar to or higher than that of the UZrH 2 - x alloy, TRIGA fuel, at temperatures from 773 to 1073 K and under hydrogen pressures from 10 2 to 10 5 Pa. Regarding the microstructure, the alloy hydrides consisted of three phases; α-U, ZrH 2 - x and ThZr 2H 7 - x , which are finely and homogeneously mixed with each other probably because they were formed from one solid solution phase stable at high temperatures. In the case of Th-rich alloys the α-U was about 1 μm in diameter and dispersed in the bulk of ThZr 2 H 7 - x and ZrH 2 - x. Such microstructure is quite similar to that of UZrH 2 - x fuel. The hydrides irradiated to 7.4 × 10 23 n/m 2 in the reactor showed a high phase stability.

  12. 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.

  13. Irradiation damage from low-dose high-energy protons on mechanical properties and positron annihilation lifetimes of Fe-9Cr alloy

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Fukumoto, K.; Ishi, Y.; Kuriyama, Y.; Uesugi, T.; Sato, K.; Mori, Y.; Yoshiie, T.

    2016-01-01

    Nuclear reactions in accelerator-driven systems (ADS) result in the generation of helium within the ADS materials. The amount of helium produced in this way is approximately one order of magnitude higher than that generated by nuclear fusion. As helium is well-known to induce degradation in the mechanical properties of metals, its effect on ADS materials is an important factor to assess. The results obtained in this study show that low-dose proton irradiation (11 MeV at 573 K to 9.0 × 10-4 dpa and 150 MeV at room temperature to 2.6 × 10-6 dpa) leads to a decrease in yield stress and ultimate tensile strength in a Fe-9Cr alloy. Moreover, interstitial helium and hydrogen atoms, as well as the annihilation of dislocation jogs, were identified as key factors that determine the observed softening of the alloy.

  14. 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.

  15. 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.

  16. Atomic kinetic Monte Carlo model based on ab initio data: Simulation of microstructural evolution under irradiation of dilute Fe CuNiMnSi alloys

    NASA Astrophysics Data System (ADS)

    Vincent, E.; Becquart, C. S.; Domain, C.

    2007-02-01

    The embrittlement of pressure vessel steels under radiation has been long ago correlated with the presence of Cu solutes. Other solutes such as Ni, Mn and Si are now suspected to contribute also to the embrittlement. The interactions of these solutes with radiation induced point defects thus need to be characterized properly in order to understand the elementary mechanisms behind the formation of the clusters formed upon radiation. Ab initio calculations based on the density functional theory have been performed to determine the interactions of point defects with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si) in order to build a database used to parameterise an atomic kinetic Monte Carlo model. Some results of irradiation damage in dilute Fe-CuNiMnSi alloys obtained with this model are presented.

  17. 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.

  18. Voids in fast-neutron-irradiated Cu, Ni and Cu-Ni concentrated alloys studied by TEM and positron annihilation methods

    NASA Astrophysics Data System (ADS)

    Fukushima, H.; Ochiai, K.; Shimomura, Y.

    The effect of concentrated Ni and Cu solute atoms in the Cu-Ni system on the formation of voids has been examined using Cu, Cu-8 at.% Ni, Ni-8 at.% Cu and Ni irradiated with fast-neutrons in the FFTF-MOTA. Both solute atoms introduced smaller voids in the grains of the concentrated alloys than voids in the normal grains of pure-Cu and pure-Ni. Slight increase of irradiation temperature and the higher dose of fast-neutrons induced coalescence of voids in the grains of Ni-8 at.% Cu, but it resulted in the abrupt decrease of the concentration of small voids in the grains and the formation of heterogeneously distributed larger voids near grain boundaries in Cu-8 at.% Ni. Heterogeneous distribution of larger voids was also observed in other materials. Annealing at higher temperatures induced segregation of impurity atoms at a void surface in Ni-8 at.% Cu.

  19. Iron redistribution in a zirconium alloy after neutron and proton irradiation studied by energy-dispersive X-ray spectroscopy (EDX) using an aberration-corrected (scanning) transmission electron microscope

    NASA Astrophysics Data System (ADS)

    Francis, E. M.; Harte, A.; Frankel, P.; Haigh, S. J.; Jädernäs, D.; Romero, J.; Hallstadius, L.; Preuss, M.

    2014-11-01

    Zirconium alloys used as cladding materials in nuclear reactors can exhibit accelerated irradiation induced growth, often termed linear growth, after sustained neutron irradiation. This phenomenon has been linked to the formation of -component dislocation loops and to the concentration of interstitial solute atoms. It is well documented for the Zircaloys that Fe dissolves from second phase particles (SPPs) during irradiation thus increasing the interstitial solute concentration in the matrix. However, no progress has yet been made into understanding whether a similar process occurs for the newer ZIRLO™ alloys. We aim to overcome this shortcoming here by studying compositional changes in second phase particles in Low Tin ZIRLO™ after neutron and proton irradiation using energy dispersive X-ray (EDX) spectroscopy. Material irradiated to 18 dpa (displacements per atom) using neutrons and to 2.3 and 7 dpa by protons was investigated. The results show that Fe is lost from Zr-Nb-Fe-SPPs during both neutron and proton irradiation. Prior to irradiation, Fe was detected at the interface of β-Nb-SPPs. This Fe enrichment is also dispersed during irradiation. Qualitatively, excellent agreement was found regarding the elemental redistribution processes observed after proton and neutron irradiation.

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

    DOE PAGESBeta

    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

  1. Tuning of the optical properties of In-rich In{sub x}Ga{sub 1−x}N (x=0.82−0.49) alloys by light-ion irradiation at low energy

    SciTech Connect

    De Luca, Marta; Polimeni, Antonio; Capizzi, Mario; Pettinari, Giorgio; Ciatto, Gianluca; Fonda, Emiliano; Amidani, Lucia; Boscherini, Federico; Knübel, Andreas; Cimalla, Volker; Ambacher, Oliver; Giubertoni, Damiano; Bersani, Massimo

    2013-12-04

    The effects of low-energy irradiation by light ions (H and He) on the properties of In-rich In{sub x}Ga{sub 1−x}N alloys are investigated by optical and structural techniques. H-irradiation gives rise to a remarkable blue-shift of light emission and absorption edge energies. X-ray absorption measurements and first-principle calculations address the microscopic origin of these effects.

  2. Small punch test evaluation of neutron-irradiation-induced embrittlement of a Cr-Mo low-alloy steel

    SciTech Connect

    Song, S.-H. . E-mail: shsonguk@yahoo.co.uk; Faulkner, R.G.; Flewitt, P.E.J.; Marmy, P.; Weng, L.-Q.

    2004-09-15

    Neutron-irradiation-induced embrittlement of a 2.25Cr1Mo steel is investigated by means of small punch testing along with scanning electron microscopy. There is an apparent irradiation-induced embrittlement effect after the steel is irradiated at about 400 deg. C for 86 days with a neutron dose rate of 1.75x10{sup -8} dpa/s. The embrittlement is mainly nonhardening embrittlement caused by impurity grain boundary segregation.

  3. A dual ion irradiation study of helium-dpa interactions on cavity evolution in tempered martensitic steels and nanostructured ferritic alloys

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takuya; Wu, Yuan; Robert Odette, G.; Yabuuchi, Kiyohiro; Kondo, Sosuke; Kimura, Akihiko

    2014-06-01

    Cavity evolutions in a normalized and tempered martensitic steel (TMS) and two nanostructured ferritic alloys (NFA) under Fe3+ and He+ dual ion beam irradiations (DII) at 500 °C and 650 °C were characterized. The irradiation conditions encompass a wide range of displacement per atom damage (dpa), He and He/dpa. The 500 °C DII produced damage and He levels of ≈10-47 dpa and ≈400-2000 appm, respectively. Transmission electron microscopy (TEM) showed that DII of a 8Cr TMS, at 500 °C to up to 60 dpa and 2100 appm He, produced a moderate density of non-uniformly distributed cavities with bimodal sizes ranging from ≈1 nm He bubbles to ≈20 nm faceted voids, and swelling ≈0.44%. In contrast, the same irradiation conditions produced only small ≈1.3 nm diameter bubbles and swelling of ≈0.05% in the NFA MA957. Similar bubble distributions were observed in MA957 and a developmental NFA DII at 650 °C up to ≈80 dpa and ≈3900 appm He. These results demonstrate the outstanding He management capability of the oxide nano-features in the NFA. The various data trends are shown as a function of dpa, He, He/dpa and He*dpa.

  4. Radiation damage of F/M and ODS alloys after Fe3+-ion irradiation at 300 °C

    NASA Astrophysics Data System (ADS)

    Kang, Suk Hoon; Chun, Young-Bum; Noh, Sanghoon; Jang, Jinsung; Jeong, Yong-Hwan; Kim, Tae Kyu

    2015-02-01

    In this study, Fe3+ self-ion irradiation is used as means of introducing irradiation damage in ferritic/martensitic (F/M) steel and oxide dispersion strengthened (ODS) steel. The ion accelerator named DuET (at Kyoto University, Japan) was used for irradiation with 6.4 MeV Fe3+ ions at 300 °C. The total number of accelerated ions was 2.5 × 1020 ions/m2, and the maximum damage rates in the F/M and the ODS steels were estimated to be roughly 6 dpa. The irradiation-induced hardness change in the damaged layer was evaluated by using nano-indentation. The F/M steel and the ODS steel commonly exhibited irradiation hardening; however, the irradiation hardening was more active in the F/M steel than in the ODS steel. The microstructure evolutions after the irradiation were investigated; point or line defects were dominantly observed in the F/M steel, while small circular cavities were typically observed in ODS steel.

  5. Irradiation creep and swelling from 400 to 600 °C of the oxide dispersion strengthened ferritic alloy MA957

    NASA Astrophysics Data System (ADS)

    Toloczko, M. B.; Gelles, D. S.; Garner, F. A.; Kurtz, R. J.; Abe, K.

    2004-08-01

    An irradiation creep and swelling study was performed on the Y 2O 3-strengthened ODS ferritic steel MA957. Pressurized tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 to 110 dpa at temperatures ranging from 400 to 600 °C. None of the stress-free tubes exhibited any evidence of swelling as determined by diameter change measurements. With a few exceptions, the irradiation creep behavior is similar to that of conventional ferritic-martensitic steels. Calculated creep compliance values are equal to those of HT9 irradiated within the same temperature range, except at 600 °C where the creep rate of MA957 is about one-half the value for HT9. The magnitude of the creep transient for MA957 is comparable to HT9, again except at 600 °C where the transient is much lower for MA957.

  6. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  7. 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.

  8. 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).

  9. 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

  10. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

    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 Es was varied from 1 to 5 J/cm2, 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.

  12. Effect of Fe and Zr ion implantation and high-current electron irradiation treatment on chemical and mechanical properties of Ti-V-Al Alloy

    NASA Astrophysics Data System (ADS)

    Pogrebnjak, Alexander D.; Kobzev, Alexander P.; Gritsenko, Boris P.; Sokolov, Sergey; Bazyl, Elena; Sviridenko, Nikolai V.; Valyaev, Alexander N.; Ivanov, Yurii F.

    2000-03-01

    Using Rutherford backscattering spectroscopy, nuclear elastic resonance analysis, atomic force microscopy, transmission electron microscopy, and wear resistance and microhardness tests, the alloy Ti41-V41-Al18 was investigated after Fe ion (60 kV) and Zr (40 kV) ion implantation and subsequent high-current electron beam (HCEB) irradiation at an energy flow density of 6 J/cm2, called duplex treatment. Profiles show that the maximum concentration of Fe ions was 16.5 at. % at 85 nm from sample surface and that of Zr ions was 0.85 at. % at 56 nm. The maximum of the Fe concentration profile was found to shift to the large sample depth toward after increasing the implantation dose. The surface alloy layer is composed of a number of structures: grains of dislocation substructure (2×1010 cm-2), grains with plates, and grains with packed martensite. The disorientation of regions is observed. After HCEB treatment, the disorientation of microregions increases (Δα=7.5°) and particles of Ti2Fe are formed. After double implantation, the dry friction wear and the friction coefficient decreased. After the duplex treatment, the thickness of the hardened layer and the wear resistance increased.

  13. Effects of heavy-ion irradiation on microstructure of V-4Cr-4Ti alloy at moderate temperatures

    SciTech Connect

    Gazda, J. |; Chung, H.M.; Loomis, B.A.; Meshii, M.

    1996-12-01

    V-4Cr-4Ti is promising for first-wall and structural applications in magnetic fusion reactors. Fast neutron sources were used to evaluate postirradiation properties of fusion candidate materials, but FFTF and EBR-II have been shutdown. Under these circumstances, heavy-ion irradiation is an attractive alternative. We used 3-MeV V{sup +} and 4.5-MeV Ni{sup 2+} ions to investigate effects of irradiation on microstructure of V-4Cr-4Ti at 200-420 C. This paper report results of ion irradiation experiments and compare with fast-neutron irradiation data. From TEM, the dominant feature of postirradiation microstructure was a high density of dislocation loops and point- defect clusters. Density and defect size depend on irradiation dose and temperature. Precipitates and voids/bubbles were not observed, even in specimens simultaneous injected with he and exposed to heavy ions. Increased transport of point defects to internal interfaces was observed, as manifested by defect denuded zones along grain boundaries. Defect denuded zones along grain boundaries could lead to segregation of impurities and solutes and formation of precipitates on grain boundaries.

  14. Comparative small-angle neutron scattering study of neutron-irradiated Fe, Fe-based alloys and a pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Bergner, F.; Lambrecht, M.; Ulbricht, A.; Almazouzi, A.

    2010-04-01

    Irradiation-induced damage in reactor pressure vessel steels covers a multitude of different features at the nanometer size scale. The nature, formation kinetics and relative importance of these features are not yet well understood in detail. It also turned out that there is no single experimental technique capable of closing all the remaining gaps. The present approach is based on the idea that significant progress can be achieved by investigating the same set of neutron-irradiated model alloys of increasing complexity with several experimental techniques including transmission electron microscopy (TEM), atom probe tomography (APT), positron annihilation spectroscopy (PAS) and small-angle neutron scattering (SANS). The aim of the effort is to explore both complementarity and overlaps of the information gained from individual techniques and to close gaps by introducing proper models. In the present paper the results obtained by means of SANS are reported, self-consistent interpretation is given and the results are qualified for the discussion in combination with the other experimental techniques to be given in separate papers.

  15. Role of nickel and manganese in recovery of resistivity in iron-based alloys after low-temperature proton irradiation

    NASA Astrophysics Data System (ADS)

    Murakami, K.; Iwai, T.; Abe, H.; Sekimura, N.; Katano, Y.; Iwata, T.; Onitsuka, T.

    2015-05-01

    This study investigates the recovery of electric resistivity in pure iron, Fe-0.6Ni and Fe-1.5Mn as related to isochronal annealing following 1 MeV proton irradiation at lower temperature than 70 K, focusing on the relationship between solute atoms and irradiation defects. Both nickel and manganese prevent stage ID recovery, which corresponds to correlated recombination. Stage II recovery is also changed by the addition of a solute, which corresponds to the migration of small interstitial clusters. In both pure iron and Fe-0.6Ni, no evident difference was observed in the stage III region, which corresponds to the migration of vacancies. In contrast, two substages appeared in the Fe-1.5Mn at a higher temperature than stage IIIB appeared in pure iron. These substages are considered to represent the release of irradiation-induced defects, which was trapped by manganese.

  16. 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 PAGESBeta

    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

  17. 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

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Saleh, T. A.; Anderoglu, O.; Romero, T. J.; Odette, G. R.; Yamamoto, T.; Li, S.; Cole, J. I.; Fielding, R.

    2016-01-01

    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(ɛ) laws for all of these alloys. In the irradiated condition, the σ(ɛ) 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 the latter can be understood in terms of the alloy's σ(ɛ) behavior. Increases in the average σ(ɛ) in the range of 0-10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also 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. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  18. 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.

  19. Irradiation Creep and Swelling from 400 C to 600 C of the Oxide Dispersion Strengthened Ferritic Alloy MA957

    SciTech Connect

    Toloczko, Mychailo B.; Gelles, David S.; Garner, Francis A.; Kurtz, Richard J.; Abe, Katsunori

    2004-04-24

    Recently, there has been a growing interest in the use of oxide dispersion strengthened (ODS) ferritic steels for fusion reactor applications. As part of an extensive study performed at PNNL on the ODS steel MA957 [1], irradiation creep tests were performed on pressurized tubes made from MA957 by two different methods. The tubes were made either by gun drilling alone or by a combination of rod drawing and gun drilling. The different fabrication methods were explored because ODS steels have been difficult to form. The pressurized tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 dpa to 110 dpa at temperatures ranging from 400 C to 600 C. The effective stresses resulting from the pressurization of the tubes ranged from 0 MPa to 175 MPa.

  20. Annealing characteristics of amorphous silicon alloy solar cells irradiated with 1.00 MeV protons

    NASA Technical Reports Server (NTRS)

    Abdulaziz, Salman S.; Woodyard, James R.

    1991-01-01

    Amorphous Si:H and amorphous Si sub x, Ge sub (1-x):H solar cells were irradiated with 1.00 MeV proton fluences in the range of 1.00E14 to 1.25E15 cm (exp -2). Annealing of the short circuit current density was studied at 0, 22, 50, 100, and 150 C. Annealing times ranged from an hour to several days. The measurements confirmed that annealing occurs at 0 C and the initial characteristics of the cells are restored by annealing at 200 C. The rate of annealing does not appear to follow a simple nth order reaction rate model. Calculations of the short-circuit current density using quantum efficiency measurements and the standard AM1.5 global spectrum compare favorably with measured values. It is proposed that the degradation in J sub sc with irradiation is due to carrier recombination through the fraction of D (o) states bounded by the quasi-Fermi energies. The time dependence of the rate of annealing of J sub sc does appear to be consistent with the interpretation that there is a thermally activated dispersive transport mechanism which leads to the passivation of the irradiation induced defects.

  1. Microstructural evolution under high flux irradiation of dilute Fe CuNiMnSi alloys studied by an atomic kinetic Monte Carlo model accounting for both vacancies and self interstitials

    NASA Astrophysics Data System (ADS)

    Vincent, E.; Becquart, C. S.; Domain, C.

    2008-12-01

    Under neutron irradiation, a large amount of point defects (vacancies and interstitials) are created. In the irradiated pressure vessel steels, weakly alloyed, these point defects are responsible for the diffusion of the solute atoms, leading to the formation of solute rich precipitates within the matrix. Ab initio calculations based on the density functional theory have been performed to determine the interactions of point defects with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si). For Mn, the results of these calculations lead to think that solute transport in α-Fe can very likely take place through an interstitial mechanism as well as via vacancies while the other solutes (Cu, Ni and Si) which establish strong bonds with vacancies diffuse more likely via vacancies only. The database thus created has been used to parameterize an atomic kinetic Monte Carlo model taking into account both vacancies and interstitials. Some results of irradiation damage in dilute Fe-CuNiMnSi alloys obtained with this model will be presented.

  2. The dependence of helium generation rate on nickel content of Fe Cr Ni alloys irradiated to high dpa levels in EBR-II

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    Fusion-relevant helium-effects experiments conducted on austenitic steels in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF) fast reactor had to recognize the contributions of both the high neutron energy (n,α) reactions and that of the 58Ni(n,γ) 59Ni(n,α) reaction sequence with low energy neutrons. An experiment conducted in the harder neutron spectra found within the core of Experimental Breeder Reactor-II (EBR-II) has shown that the helium in this reactor was generated almost exclusively from the interaction of high energy neutrons with the natural isotopes of nickel. There was very little contribution from 59Ni. The helium production was found to scale directly with the nickel content over the range 25-75% Ni. Even at very high neutron exposures, the helium production in such reactors can be predicted within 5% accuracy on the basis of high energy reactions, as demonstrated by an experiment conducted on three Fe-15Cr-Ni ternary alloys irradiated to doses of 75-131 dpa in EBR-II.

  3. 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.

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

    PubMed

    Parish, Chad M; Miller, Michael K

    2014-04-01

    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, that are 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 cavities such as helium bubbles from the Ti-Y-O rich nanoclusters (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. Fresnel-contrast imaging, with careful attention to TEM-STEM reciprocity, differentiates bubbles from NCs. 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. PMID:24598435

  5. A ternary EAM interatomic potential for U-Mo alloys with xenon

    NASA Astrophysics Data System (ADS)

    Smirnova, D. E.; Kuksin, A. Yu; Starikov, S. V.; Stegailov, V. V.; Insepov, Z.; Rest, J.; Yacout, A. M.

    2013-04-01

    A new interatomic potential for a uranium-molybdenum system with xenon is developed in the framework of an embedded atom model using a force-matching technique and a dataset of ab initio atomic forces. The verification of the potential proves that it is suitable for the investigation of various compounds existing in the system as well as for simulation of pure elements: U, Mo and Xe. Computed lattice constants, thermal expansion coefficients, elastic properties and melting temperatures of U, Mo and Xe are consistent with the experimentally measured values. The energies of the point defect formation in pure U and Mo are proved to be comparable to the density-functional theory calculations. We compare this new U-Mo-Xe potential with the previously developed U and Mo-Xe potentials. A comparative study between the different potential functions is provided. The key purpose of the new model is to study the atomistic processes of defect evolution taking place in the U-Mo nuclear fuel. Here we use the potential to simulate bcc alloys containing 10 wt% of intermetallic Mo and U2Mo.

  6. 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

  7. Effect of swift heavy ion irradiation on the physical properties of CuIn(S 0.4Se 0.6) 2 alloy thin films prepared by solution growth technique

    NASA Astrophysics Data System (ADS)

    Chavhan, S. D.; Deshpande, N. G.; Gudage, Y. G.; Ghosh, A.; Ahire, R. R.; Borse, S. V.; Khairnar, R. S.; Jadhav, K. M.; Singh, F.; Sharma, Ramphal

    2008-06-01

    Alloy thin films of CuIn(S 0.4Se 0.6) 2 material were deposited using the solution growth technique. The various deposition parameters such as pH of solution, time, concentration of ions and temperature have been optimized for the device grade thin films. The as-deposited films were annealed in a rapid thermal annealing (RTA) system at 450 °C in air for 5 min and subjected to high-energy Ag ion irradiations. Ag ion irradiation has been performed with an energy of 100 MeV at a fluency of 5×10 12 ions/cm 2 on the thin film. The changes in optical and electrical properties that occurred before and after post-deposition treatments in CuIn(S 0.4Se 0.6) 2 thin films were studied using X-ray diffraction (XRD) and AFM; increase in crystallinity was observed after annealing and irradiation. In addition, structural damages were observed in irradiated thin films. After annealing and irradiation, the surface roughness was seen to be increased. Decrease in resistivity was observed, which is consistent with the optical energy band gap. The results are explained by considering the high energy deposited due to the electronic energy loss upon irradiation, which modified the properties of the material.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Hydrogen in titanium alloys

    SciTech Connect

    Wille, G W; Davis, J W

    1981-04-01

    The titanium alloys that offer properties worthy of consideration for fusion reactors are Ti-6Al-4V, Ti-6Al-2Sn-4Zr-2Mo-Si (Ti-6242S) and Ti-5Al-6Sn-2Zr-1Mo-Si (Ti-5621S). The Ti-6242S and Ti-5621S are being considered because of their high creep resistance at elevated temperatures of 500/sup 0/C. Also, irradiation tests on these alloys have shown irradiation creep properties comparable to 20% cold worked 316 stainless steel. These alloys would be susceptible to slow strain rate embrittlement if sufficient hydrogen concentrations are obtained. Concentrations greater than 250 to 500 wppm hydrogen and temperatures lower than 100 to 150/sup 0/C are approximate threshold conditions for detrimental effects on tensile properties. Indications are that at the elevated temperature - low hydrogen pressure conditions of the reactors, there would be negligible hydrogen embrittlement.

  14. 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.

  15. 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.

  16. 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.

  17. Combined ion (Ar+, 20 keV) and light irradiation of the quenched Fe-8.25 at % Mn alloy. Separation between thermal and radiation induced long-range effects

    NASA Astrophysics Data System (ADS)

    Ovchinnikov, V. V.; Gushchina, N. V.; Bedin, S. A.

    2016-02-01

    Mössbauer and X-ray diffraction investigations of the radiation-induced α → γ phase transformation and short-range-order formation processes in the quenched Fe-8.25 at % Mn alloy under combined exposure (simultaneous visible light and Ar+ 20-keV ion beam irradiation) are carried out. Combined exposure made it possible to fix the target stationary temperature, and hence, the intensity of thermally-stimulated processes; an energy and ion current density could independently be varied in a wide range. As a result, an important contribution of a non-thermal constituent of ion beam exposure to the structural state of alloy was proved. Only in the presence of ion beam, an α → γ (bcc → fcc) phase transformation and accelerated intraphase processes preparing this transformation are observed in the deep layers of the target (about 103 Rp). With allowance for the relatively low level of thermally and radiation-stimulated processes, radiation-dynamic effects associated with propagation of intense post-cascade solitary waves, which can rearrange metastable matters, are considered as the cause of the observed transformations.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    In this last of a series of papers describing the evolution of microstructure, void swelling and mechanical properties of model austenitic alloys in response to differences in helium/dpa rates, shear punch testing is used to assess the relative effect of helium generation ratios and various important material and environmental variables. Shear punch data confirm the general trends observed in earlier tensile data derived from the 59Ni isotopic doping experiment. There is a convergence to a common saturation level of yield strength that depends on alloy composition, temperature and displacement rate, but not on starting condition. The approach to saturation can be sensitive to helium/dpa ratio, however, and may depend on the starting state. For reasons not yet known, shear punch tests appear to be more sensitive to such transient differences than are tensile tests.

  20. 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.

  1. 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.

  2. 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.

  3. Charpy impact test results for low-activation ferritic alloys

    SciTech Connect

    Cannon, N.S.; Hu, W.L.; Gelles, D.S.

    1987-05-01

    The objective of this work is to evaluate the shift of the ductile to brittle transition temperature (DBTT) and the reduction of the upper shelf energy (USE) due to neutron irradiation of low activation ferritic alloys. Six low activation ferritic alloys have been tested following irradiation at 365/sup 0/C to 10 dpa and compared with control specimens in order to assess the effect of irradiation on Charpy impact properties.

  4. Stress corrosion cracking and intergranular corrosion of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Fukuya, K.; Shima, S.; Kayano, H.; Narui, M.

    1992-09-01

    The effects of irradiation on stress corrosion cracking (SCC) and intergranular corrosion (IGC) susceptibility were investigated in solution-treated Fe19Cr9NiMn alloys and JPCA irradiated to 5.3×1024 n/m2 (E > 1 MeV) at 573 K. In Fe19Cr9NiMn alloys, the irradiation enhanced IGC i n boiling HNO3 + Cr6+ solution when the alloys contained phosphorus and silicon and induced SCC in all the alloys with strain rate tensile tests in 571 K water containing 32 ppm oxygen. With increasing phosphorus and silicon contents. IGC was promoted but IGSCC was suppressed after irradiation. The results indicated that these elements are not the main contributors to irradiation-assisted SCC, although they affect SCC behavior. The Japanese Prime Candidate Alloy (JPCA) had better SCC resistance than Fe19Cr9NiMn alloys under the present irradiation condition.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. The effect of alloy composition on radiation-induced segregation in FeCrNi alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Was, G. S.; Kenik, E. A.

    1997-04-01

    The effect of alloy composition on radiation-induced segregation (RIS) was investigated in austenitic iron-base and nickel-base alloys using proton irradiation. Specifically, RIS was studied by irradiation of Ni18Cr, Ni18Cr9Fe, and Fe20Cr9Ni over a dose range of 0 to 1.0 dpa and a temperature range of 200 to 500°C. Grain boundary composition was measured using Auger electron spectroscopy and scanning transmission electron microscopy with energy dispersive X-ray spectroscopy. Measurements from this study along with measurements from Fe16Cr24Ni, Fe20Cr24Ni, Fe24Cr24Ni, and Fe24Cr19Ni alloys irradiated with protons confirm that RIS is strongly dependent on the alloy composition. Trends in segregation behavior in Fe-base alloys are consistent with high temperature diffusion measurements, indicating that a vacancy mechanism is the most likely primary driving force for RIS in austenitic steels. The migration energy for Cr is shown to be larger than the migration energy of Fe. Segregation measurements in Ni-base alloys are not consistent with high temperature diffusion measurements, indicating that ordering forces may be significant in the segregation process. Comparison of model calculations to measured RIS data indicate that Fe, Cr, and Ni diffusivities are composition dependent. This dependence on alloy composition limits the predictive ability of simple models because of the need for separate diffusion parameters for every alloy composition.

  10. 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.

  11. 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.

  12. 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.

  13. Manufacturing and STA-investigation of witness-samples for the temperature monitoring of structural steels under irradiation

    NASA Astrophysics Data System (ADS)

    Sevryukov, O. N.; Fedotov, V. T.; Polyansky, A. A.; Pokrovski, S. A.; Kuzmin, R. S.

    2016-04-01

    The object of investigations was alloys based on lead and cadmium used as fuse monitors to control the maximum irradiation temperature (fuse temperature monitors, FTM) of samples from structural steels under irradiation in a research reactor IR-8. The result of the work was selected and tested initial materials for production of alloys. A technological scheme of the production of alloys for FTM has been developed and experimental studies of the properties of these alloys have been carried out.

  14. 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.

  15. 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.

  16. Cleavage mechanism in vanadium alloys

    SciTech Connect

    Odette, G.R.; Donahue, E.; Lucas, G.E.

    1997-12-31

    The effect specimen geometry, loading rate and irradiation on the ductile-to-brittle transition in a V-4Ti-4Cr alloy were evaluated and modeled. Confocal microscopy-fracture reconstruction and SEM were used to characterize the sequence-of-events leading to cleavage, as well as the CTOD at fracture initiation. This alloy undergoes normal stress-controlled transgranular cleavage below a transition temperature that depends primarily on the tensile properties and constraint. Thus an equivalent yield stress model is in good agreement with observed effects of loading rate and irradiation hardening. Predicted effects of specimen geometry based on a critical stress-area criteria and FEM simulations of crack tip fields were also found to be in agreement with experiment. Some interesting characteristics of the fracture process are also described.

  17. 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))))))))

  18. Aluminum alloy

    NASA Technical Reports Server (NTRS)

    Blackburn, Linda B. (Inventor); Starke, Edgar A., Jr. (Inventor)

    1989-01-01

    This invention relates to aluminum alloys, particularly to aluminum-copper-lithium alloys containing at least about 0.1 percent by weight of indium as an essential component, which are suitable for applications in aircraft and aerospace vehicles. At least about 0.1 percent by weight of indium is added as an essential component to an alloy which precipitates a T1 phase (Al2CuLi). This addition enhances the nucleation of the precipitate T1 phase, producing a microstructure which provides excellent strength as indicated by Rockwell hardness values and confirmed by standard tensile tests.

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  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. Elevated temperature aluminum alloys

    NASA Technical Reports Server (NTRS)

    Meschter, Peter (Inventor); Lederich, Richard J. (Inventor); O'Neal, James E. (Inventor)

    1989-01-01

    Three aluminum-lithium alloys are provided for high performance aircraft structures and engines. All three alloys contain 3 wt % copper, 2 wt % lithium, 1 wt % magnesium, and 0.2 wt % zirconium. Alloy 1 has no further alloying elements. Alloy 2 has the addition of 1 wt % iron and 1 wt % nickel. Alloy 3 has the addition of 1.6 wt % chromium to the shared alloy composition of the three alloys. The balance of the three alloys, except for incidentql impurities, is aluminum. These alloys have low densities and improved strengths at temperatures up to 260.degree. C. for long periods of time.

  12. 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.

  13. 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.

  14. 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.

  15. Irradiation embrittlement of neutron-irradiated low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kayano, H.; Kimura, A.; Narui, M.; Sasaki, Y.; Suzuki, Y.; Ohta, S.

    1988-07-01

    Effects of neutron irradiation and additions of small amounts of alloying elements on the ductile-brittle transition temperature (DBTT) of three different groups of ferritic steels were investigated by means of the Charpy impact test in order to gain an insight into the development of low-activation ferritic steels suitable for the nuclear fusion reactor. The groups of ferritic steels used in this study were (1) basic 0-5% Cr ferritic steels, (2) low-activation ferritic steels which are FeCrW steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) FeCrMo, Nb or V ferritic steels for comparison. In Fe-0-15% Cr and FeCrMo steels, Fe-3-9% Cr steels showed minimum brittleness and provided good resistance against irradiation embrittlement. Investigations on the effects of additions of trace amounts of alloying elements on the fracture toughness of low-activation ferritic steels made clear the optimum amounts of each alloying element to obtain higher toughness and revealed that the 9Cr-2W-Ta-Ti-B ferritic steel showed the highest toughness. This may result from the refinement of crystal grains and improvement of quenching characteristics caused by the complex effect of Ti and B.

  16. 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.

  17. 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.

  18. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2014-01-01

    The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

  19. [Food irradiation].

    PubMed

    Migdał, W

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by Codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and the World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Institute of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19MeV, 1 kW) and an industrial unit Elektronika (10MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permission for irradiation for: spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables. PMID:8619113

  20. 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.

  1. 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.

  2. Properties of copper?stainless steel HIP joints before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Tähtinen, S.; Laukkanen, A.; Singh, B. N.; Toft, P.

    2002-12-01

    The tensile and fracture behaviour of CuCrZr and CuAl25 IG0 alloys joint to 316L(N) stainless steel by hot isostatic pressing (HIP) have been determined in unirradiated and neutron-irradiated conditions. The tensile and fracture behaviour of copper alloy HIP joint specimens are dominated by the properties of the copper alloys, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloys and stainless steel. The test temperature, neutron irradiation and thermal cycles primarily affect the copper alloy HIP joint properties through changing the strength mismatch between the base alloys. Changes in the loading conditions i.e. tensile, bend ( JI) and mixed-mode bend ( JI/ JII) lead to different fracture modes in the copper alloy HIP joint specimens.

  3. 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.

  4. 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)

  5. Irradiance gradients

    SciTech Connect

    Ward, G.J. Ecole Polytechnique Federale, Lausanne ); Heckbert, P.S. . School of Computer Science Technische Hogeschool Delft . Dept. of Technical Mathematics and Informatics)

    1992-04-01

    A new method for improving the accuracy of a diffuse interreflection calculation is introduced in a ray tracing context. The information from a hemispherical sampling of the luminous environment is interpreted in a new way to predict the change in irradiance as a function of position and surface orientation. The additional computation involved is modest and the benefit is substantial. An improved interpolation of irradiance resulting from the gradient calculation produces smoother, more accurate renderings. This result is achieved through better utilization of ray samples rather than additional samples or alternate sampling strategies. Thus, the technique is applicable to a variety of global illumination algorithms that use hemicubes or Monte Carlo sampling techniques.

  6. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

    The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

  7. 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.

  8. Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

    DOE PAGESBeta

    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

  9. Deuterium retention in VPS-W coated vanadium alloy (V-4Cr-4Ti)

    NASA Astrophysics Data System (ADS)

    Ueda, Yoshio; Uekita, Kazuhiro; Oya, Makoto; Ohtsuka, Yusuke; Nagasaka, Takuya; Kasada, Ryuta; Kimura, Akihiko; Tokunaga, Tomonori; Yoshida, Naoaki

    2013-07-01

    D retention in a VPS W coated V alloy (V-4Cr-4Ti, NIFS Heat 2) was investigated by an ion beam irradiation with a HiFIT device. Irradiation temperature changed from 473 K to 773 K and ion fluence was 1024 m-2. In addition, during ion irradiation atmospheric pressure was 0.5 mTorr. It was found that ambient gas absorption was the dominant process for D retention. Ion irradiation could enhance gas absorption probably due to surface modification (oxide layer removal) and could produce ion induced trapping sites. Although these ion induced processes can be avoided by W coating layers, D retention is still significant for VPS-W coated V alloy. D trapping could be done mainly in V-alloy and its mechanisms could be related to some intrinsic defects in V alloy and formation of titanium hydride at elevated temperature.

  10. 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.

  11. Irradiation-induced precipitation modelling of ferritic steels

    NASA Astrophysics Data System (ADS)

    Yin, You Fa; Faulkner, Roy G.; Lu, Zheng

    2009-06-01

    In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.

  12. 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.

  13. Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems - Final Report , Project 99-0280

    SciTech Connect

    Bruemmer, Stephen M.; Simonen, Edward P.; Gan, Jian; Garner, Francis A.; Gelles, David S.; Edwards, Danny J.; Andresen, Peter L.; Young, Lisa M.; Was, Gary S.; Fournier, L.; Sencer, Bulent H.

    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.

  14. 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.

  15. Small-scale characterisation of irradiated nuclear materials: Part I – Microstructure

    DOE PAGESBeta

    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

  16. 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.

  17. 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.

  18. Particle size effect on microwave absorbing of La0.67Ba0.33Mn0.94Ti0.06O3 powders prepared by mechanical alloying with the assistance of ultrasonic irradiation

    NASA Astrophysics Data System (ADS)

    Saptari, Sitti Ahmiatri; Manaf, Azwar; Kurniawan, Budhy

    2016-03-01

    Doped manganites have attracted substantial interest due to their unique chemical and physics properties, which makes it possible to be used for microwave absorbing materials. In this paper we report synthesizes and characterization of La0.67Ba0.33Mn0.94Ti0.06O3 powders prepared by mechanical alloying with the assistance of a high power ultrasonic treatment. After solid state reaction, the presence of single phase was confirmed by X-ray Diffraction (XRD). Refinement results showed that samples are single phase with monoclinic structure. It was found that powder materials derived from mechanical alloying results in large variation in the particle size. A significant improvement was obtained upon subjecting the mechanically milled powder materials to an ultrasonication treatment for a relatively short period of time. As determined by particle size analyzer (PSA), the mean particle size gradually decreased from the original size of 5.02 µm to 0.36 µm. Magnetic properties were characterized by VSM, and hysteresis loops results showed that samples are soft magnetic. It was found that when the mean particle size decreases, saturation was increases and coersitivity was decreases. Microwave absorption properties were investigated in the frequency range of 8-12 GHz using vector network analyzer. An optimal reflection loss of 24.44 dB is reached at 11.4 GHz.

  19. 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. PMID:21456241

  20. Microstructural examination of commercial ferritic alloys at 200 dpa

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1996-10-01

    Microstructures and density change measurements are reported for martensitic commercial steels HT-9 and modified 9Cr1Mo (T91) and oxide dispersion strengthened ferritic alloys MA956 and MA957 following irradiation in the FFTF/MOTA at 420°C to 200 dpa. Swelling as determined by density change remains below 2% for all conditions. Microstructures are found to be stable except in recrystallized grains of MA957, which are fabrication artifacts, with only minor swelling in the martensitic steels and α' precipitation in alloys with 12% or more chromium. These results further demonstrate the high swelling resistance and microstructural stability of the ferritic alloy class.

  1. 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)

  2. Dynamical interaction of helium bubbles with cascade damage in Fe-9Cr ferritic alloy.

    SciTech Connect

    Ono, K.; Miyamoto, M.; Arakawa, K.; Birtcher, R. C.; Materials Science Division; Shimane Univ.; Osaka Univ.

    2008-12-01

    Dynamic interaction of helium bubble with cascade damage in Fe-9Cr ferritic alloy has been studied using in situ irradiation and electron microscopy. During the irradiation of the alloy by 400 keV Fe{sup +} ions at temperatures where no thermal motion takes place, induced displacement of small helium bubbles was observed: the bubbles underwent sporadic and instant displacement. The displacement was of the order of a few nanometers. The experimentally determined displacement probability of helium bubbles is consistent with the calculated probability of their dynamic interaction with sub-cascades introduced by the irradiation. Furthermore, during the irradiation of the alloy at higher temperatures, both retarded and accelerated Brownian type motions were observed. These results are discussed on the basis of dynamic interaction of helium bubbles with point defects that survive through high-energy self-ion irradiation.

  3. 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.

  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. 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

  6. 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.

  7. Bubble formation in Zr alloys under heavy ion implantation

    SciTech Connect

    Pagano, L. Jr.; Motta, A.T.; Birtcher, R.C.

    1995-12-01

    Kr ions were used in the HVEM/Tandem facility at ANL to irradiate several Zr alloys, including Zircaloy-2 and -4, at 300-800 C to doses up to 2{times}10{sup 16}ion.cm{sup -2}. Both in-situ irradiation of thin foils as well as irradiation of bulk samples with an ion implanter were used in this study. For the thin foil irradiations, a distribution of small bubbles in the range of 30-100 {angstrom} was found at all temperatures with the exception of the Cr-rich Valloy where 130 {angstrom} bubbles were found. Irradiation of bulk samples at 700-800 C produced large faceted bubbles up to 300 {angstrom} after irradiation to 2{times}10{sup 16}ion.cm{sup -2}. Results are examined in context of existing models for bubble formation and growth in other metals.

  8. 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.

  9. Vanadium alloys - overview and recent results

    NASA Astrophysics Data System (ADS)

    Muroga, T.; Nagasaka, T.; Abe, K.; Chernov, V. M.; Matsui, H.; Smith, D. L.; Xu, Z.-Y.; Zinkle, S. J.

    2002-12-01

    This paper reviews recent progress in research on vanadium alloys with emphasis on V-4Cr-4Ti as a reference composition. New high purity V-4Cr-4Ti ingots and products (NIFS-HEATs) were made. The improved purity of the alloys made a practical demonstration of enhanced feasibility of recycling as a method of handling after use in fusion reactors. Significant progress has been made in the understanding of physical metallurgy of V-4Cr-4Ti and effects of O, N and C on the alloy properties such as low and high temperature mechanical properties, welding properties and low temperature irradiation effects, by means of including the comparison of various large heats and model alloys with different impurity levels. The effects of other trace impurities on some of the properties are also discussed. Other current efforts to characterize V-4Cr-4Ti, to improve its properties and to explore advanced vanadium alloys are reviewed. Issues remaining for the future investigations are discussed.

  10. 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.

  11. 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.

  12. 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) .

  13. 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.

  14. 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.

  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. 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.

  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. 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/.

  19. 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.

  20. Cross-sectional TEM studies of DIGM in irradiated Au-Cu bilayers

    SciTech Connect

    Gao, Y.; Alexander, D.E.; Rehn, L.E.

    1992-11-01

    Cross-sectional transmission electron microscopy was used to study diffusion-induced gain boundary migration (DIGM) in irradiated and annealed Au/Cu bilayers. Using this technique, in combination with small probe x-ray energy dispersive spectroscopy, DIGM alloyed zones in Au were identified in an irradiated sample.

  1. 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.

  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. 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.

  4. 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.

  5. High strength alloys

    SciTech Connect

    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. High strength alloys

    SciTech Connect

    Maziasz, Phillip James; Shingledecker, John Paul; Santella, Michael Leonard; Schneibel, Joachim Hugo; Sikka, Vinod Kumar; Vinegar, Harold J; John, Randy Carl; Kim, Dong Sub

    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.

  7. Effects of solid transmutants and helium in copper studied by mixed-spectrum neutron irradiation

    NASA Astrophysics Data System (ADS)

    Muroga, T.; Watanabe, H.; Yoshida, N.

    1998-10-01

    Microstructures of pure Cu and Cu-Ni-Zn alloys irradiated in High Flux Isotope Reactor (HFIR) at 573 K to 9.2 dpa and 673 K to 10.4 dpa have been observed with TEM. Transmutant Ni and Zn of ˜3 wt% were produced during the irradiation. The effect of Ni and Zn production during irradiation was interpreted based on the knowledge obtained from Fast Flux Test Facility (FFTF) irradiated Cu and Cu-Ni-Zn alloys. The effect of He produced from Ni in Cu-5Ni during irradiation in HFIR was consistent with that produced from 10B in Cu-5Ni- 10B during irradiation in FFTF.

  8. Spark alloying of an AL9 alloy by hard alloys

    NASA Astrophysics Data System (ADS)

    Kuptsov, S. G.; Fominykh, M. V.; Mukhinov, D. V.; Magomedova, R. S.; Nikonenko, E. A.

    2015-08-01

    The phase compositions of spark coatings of Kh12M steel with a VT1-0 (titanium) alloy and T15K6 and T30K4 hard alloys are studied. It is shown that the TiC titanium carbide forms in all cases and tungsten carbide decomposes with the formation of tungsten in a coating. These processes are intensified by increasing time, capacitance, and frequency. The surface hardness, the sample weight, and the white layer thickness increase monotonically.

  9. 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.

  10. 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

  11. Influence of irradiation on KISCC of Zr-1%Nb claddings

    NASA Astrophysics Data System (ADS)

    Bibilashvili, Yu. K.; Medvedev, A. V.; Nesterov, B. I.; Novikov, V. V.; Golovanov, V. N.; Eremin, S. G.; Yurtchenko, A. D.

    2000-06-01

    Experimental results on iodine induced stress corrosion cracking (SCC) in irradiated claddings from Zr-1%Nb alloy are analyzed. Fatigue cracks were grown at their inner surfaces. The irradiation was carried on in the liquid sodium cooled BOR-60 to the fluence not lower than 10 22 n/cm 2. The SCC-test was carried at 350°C, in argon gas at constant pressure and iodine surface concentration of 0.2 mg/cm 2. The threshold stress intensity factor for the irradiated Zr-1%Nb claddings was determined to be ˜2.0 MPa√m.

  12. The COBRA-1B irradiation experiment in EBR-II

    SciTech Connect

    Tsai, H.; Hins, A.G.; Strain, R.V.; Smith, D.L.

    1994-09-01

    The objective of the forthcoming COBRA-1B experiment in EBR-II is to evaluate the effects of fast neutron irradiation on the physical and mechanical properties of candidate fusion structural materials. Of special interest in this experiment will be ITER-relevant temperature and exposure for the test specimens. Approximately 50% of the irradiation test volume will be devoted to vanadium-alloy specimens. Design of the COBRA-1B irradiation experiment began in this reporting period and is in progress. The target reactor insertion date for COBRA-1B is September 1994. Technical and programmatic feasibility approval for the experiment has been granted by EBR-II Operations.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Observation and possible mechanism of irradiation induced creep in ceramics

    NASA Astrophysics Data System (ADS)

    Katoh, Yutai; Snead, Lance L.; Parish, Chad M.; Hinoki, Tatsuya

    2013-03-01

    Stress relaxation of elastically strained silicon carbide samples during high flux neutron irradiation to ˜2 displacements per atom at intermediate (390-540 °C) to high (790-1180 °C) temperatures is presented. The magnitude of stress relaxation normalized to the initial stress magnitude is independent of the initial stress magnitude, indicating a stress exponent of unity for irradiation creep in SiC. The creep strain increases with the increasing fluence while the strain rate significantly decreases. A linear relationship was found between the creep strain and the transient swelling that occurs due to irradiation defect accumulation. The apparent irradiation creep compliances for silicon carbide are substantially smaller than those associated with pure metals and alloys. Microstructural examination suggests that incoherent grain boundaries likely play a major role in determining the primary transient irradiation creep of these materials at high temperatures with a potential additional contribution from basal slip at very high temperatures.

  18. AFC-1 Transmutation Fuels Post-Irradiation Hot Cell Examination 4-8 at.% - Final Report (Irradiation Experiments AFC-1B, -1F and -1Æ)

    SciTech Connect

    Bruce Hilton; Douglas Porter; Steven Hayes

    2006-09-01

    The AFC-1B, AFC-1F and AFC-1Æ irradiation tests are part of a series of test irradiations designed to evaluate the feasibility of the use of actinide bearing fuel forms in advanced fuel cycles for the transmutation of transuranic elements from nuclear waste. The tests were irradiated in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) to an intermediate burnup of 4 to 8 at% (2.7 - 6.8 x 1020 fiss/cm3). The tests contain metallic and nitride fuel forms with non-fertile (i.e., no uranium) and low-fertile (i.e., uranium bearing) compositions. Results of postirradiation hot cell examinations of AFC-1 irradiation tests are reported for eleven metallic alloy transmutation fuel rodlets and five nitride transmutation fuel rodlets. Non-destructive examinations included visual examination, dimensional inspection, gamma scan analysis, and neutron radiography. Detailed examinations, including fission gas puncture and analysis, metallography / ceramography and isotopics and burnup analyses, were performed on five metallic alloy and three nitride transmutation fuels. Fuel performance of both metallic alloy and nitride fuel forms was best correlated with fission density as a burnup metric rather than at.% depletion. The actinide bearing transmutation metallic alloy compositions exhibit irradiation performance very similar to U-xPu-10Zr fuel at equivalent fission densities. The irradiation performance of nitride transmutation fuels was comparable to limited data published on mixed nitride systems.

  19. Towards Radiation Tolerant Nanostructured Ferritic Alloys

    SciTech Connect

    Miller, Michael K; Hoelzer, David T; Russell, Kaye F

    2010-01-01

    The high temperature and irradiation response of a new class of nanostructured ferritic alloys have been investigated by atom probe tomography. These materials are candidate materials for use in the extreme environments that will be present in the next generation of power generating systems. Atom probe tomography has revealed that the yttria powder is forced into solid solution during the mechanical alloying process andsubsequently 2-nm-diameter Ti-, Y- and O-enriched nanoclusters are formedduring the extrusion process. These nanoclusters have been shown to be remarkably stable during isothermal annealing treatments up to 0.92 of the melting temperature and during proton irradiation up to 3 displacements per atom. No significant difference in sizes, compositions and number densities of the nanoclusters was also observed between the unirradiated and proton irradiated conditions. The grain boundaries were found to have high number densities of nanoclusters as well as chromium and tungsten segregation which pin the grain boundary to minimize creep and grain growth.

  20. 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.

  1. Assessment of Titanium Aluminide Alloys for High-Temperature Nuclear Structural Applications

    NASA Astrophysics Data System (ADS)

    Zhu, Hanliang; Wei, Tao; Carr, David; Harrison, Robert; Edwards, Lyndon; Hoffelner, Wolfgang; Seo, Dongyi; Maruyama, Kouichi

    2012-12-01

    Titanium aluminide (TiAl) alloys exhibit high specific strength, low density, good oxidation, corrosion, and creep resistance at elevated temperatures, making them good candidate materials for aerospace and automotive applications. TiAl alloys also show excellent radiation resistance and low neutron activation, and they can be developed to have various microstructures, allowing different combinations of properties for various extreme environments. Hence, TiAl alloys may be used in advanced nuclear systems as high-temperature structural materials. Moreover, TiAl alloys are good materials to be used for fundamental studies on microstructural effects on irradiation behavior of advanced nuclear structural materials. This article reviews the microstructure, creep, radiation, and oxidation properties of TiAl alloys in comparison with other nuclear structural materials to assess the potential of TiAl alloys as candidate structural materials for future nuclear applications.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Void swelling in binary Fe sbnd Cr alloys at 200 dpa

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1995-08-01

    Microstructural examinations have been performed on a series of binary Fe sbnd Cr alloys irradiated in the FFTF/MOTA at 425°C to 200 dpa. The data represent the highest swelling levels reported to date in neutron-irradiated ferritic alloys. The alloy compositions ranged from 3 to 18% Cr in 3% Cr increments and the irradiation temperature corresponded to the peak swelling condition for this alloy class. Density measurements showed swelling levels as high as 7.4%, with the highest swelling found in the Fe sbnd 9Cr and sbnd 6Cr alloys. Microstructural examinations revealed that the highest swelling conditions contained well-developed voids, often as large as 100 nm, and a dislocation network comprised of both a/2<111> and a<100> Burgers vectors. Swelling was lower in the other alloys, and the swelling reduction could be correlated with increased precipitation. These results are considered in light of the current theories for low swelling in ferritic alloys, but no theory is found to completely explain the results.

  8. 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.

  9. 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.

  10. Rechargeable sodium alloy anode

    SciTech Connect

    Jow, T.R.

    1988-06-28

    A secondary battery is described comprising: (a) an anode which comprises an alloy of sodium and one or metals selected from the group consisting of tin, lead antimony, bismuth, selenium and tellerium, (b) an electrolyte comprising one or more organic solvents and one or more sodium salts dissolved therein forming dissolved sodium cations in solution; and (c) a cathode; the sodium cations from the electrolyte alloying with the one or more metals of the alloy in the anode during the charging of the battery and sodium in the alloy disoloving in the electrolyte during the discharging of the battery.

  11. Emulation of reactor irradiation damage using ion beams

    DOE PAGESBeta

    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

  12. 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.

  13. Kinetics of radiation-induced segregation in ternary alloys. [LMFBR

    SciTech Connect

    Lam, N.Q.; Kumar, A.; Wiedersich, H.

    1982-01-01

    Model calculations of radiation-induced segregation in ternary alloys have been performed, using a simple theory. The theoretical model describes the coupling between the fluxes of radiation-induced defects and alloying elements in an alloy A-B-C by partitioning the defect fluxes into those occurring via A-, B-, and C-atoms, and the atom fluxes into those taking place via vacancies and interstitials. The defect and atom fluxes can be expressed in terms of concentrations and concentration gradients of all the species present. With reasonable simplifications, the radiation-induced segregation problem can be cast into a system of four coupled partial-differential equations, which can be solved numerically for appropriate initial and boundary conditions. Model calculations have been performed for ternary solid solutions intended to be representative of Fe-Cr-Ni and Ni-Al-Si alloys under various irradiation conditions. The dependence of segregation on both the alloy properties and the irradiation variables, e.g., temperature and displacement rate, was calculated. The sample calculations are in good qualitative agreement with the general trends of radiation-induced segregation observed experimentally.

  14. Modelling Thermodynamics of Alloys for Fusion Application

    SciTech Connect

    Caro, A; Erhart, P; de Caro, M S; Sadigh, B; Srinivasan, S G; Stukowski, A

    2009-07-29

    This research has two main objectives: (1) The development of computational tools to evaluate alloy properties, using the information contained in thermodynamic functions. We aim at improving the ability of classical potentials to account for complex alloy behavior, and (2) The application of these tools to predict properties of alloys under irradiation, in particular the FeCr system. This semester has been very productive in the developments of both tools and algorithms. Our work aims at developing theoretical and numerical methodologies that are directly applicable to multi-scale modeling addressing the specific issues related to multi-component, multi-phase systems in non-equilibrium states, such as solid-solution hardening, point defect-solute interactions, stoichiometry effects, static and dynamic strain aging, dislocation-solute interactions, and in general the aspects of microstructure evolution that are affected by irradiation. At its present stage of development, we have been able to predict numerous thermodynamic properties of FeCr mainly related to ordering and precipitation; we have found new intermetallic phases and suggested the existence of a dependence of the solubility limit on the degree of order of the alloy. At present, we are studying dislocation mobility in the solid solution and the heterogeneous phase, and we are developing a new algorithm to perform Monte Carlo simulations inside the miscibility gap, a technique that will allow us to study interfacial energies and nucleation sizes. We develop a strategy to model radiation damage in FeCr alloys, system in which magnetism introduces an anomaly in the heat of formation of the solid solution that is at the basis of its unique behavior. Magnetism has implications for the precipitation of excess Cr in the a phase in the presence of heterogeneities. These complexities pose many challenges for atomistic (empirical) methods. To address such issues we develop a modified, many-body potential by

  15. Study on Defects in H+ion Implanted B2 type Fe-Al Alloy using Slow Positron Beam

    NASA Astrophysics Data System (ADS)

    Komagata, S.; Kawasuso, A.; Yabuuchi, A.; Maekawa, M.; Batchulun, C.; Yasuda, K.; Ishigami, R.; Kume, K.; Iwase, A.; Hori, F.

    Fe48-at.%Al alloy were implanted with 50 keV H+ ions to the fluence of 3×1016 and 1×1018/cm2 at room temperature. Positron annihilation Doppler broadening and lifetime measurements for these alloys have been carried out using slow positron beam apparatus with an energy range of 0.2 to 30.2 keV. The positron annihilation S-parameter decreased by H+ ion irradiation. Also the positron lifetimes for hydrogen deposited region in the alloy decreased by the irradiation. These results show that implanted H atoms were trapped by vacancy type defects.

  16. Helium entrapment in a nanostructured ferritic alloy

    SciTech Connect

    Edmondson, Philip D; Parish, Chad M; Zhang, Yanwen; Hallen, Dr Anders; Miller, Michael K

    2011-01-01

    The nanostructured ferritic alloy 14YWT has been irradiated with He ions to simulate accumulation of He during the service life of a nuclear reactor to test the hypothesis that the large surface area for nanoclusters is a preferential nucleation site for bubbles. Transmission electron microscopy and atom probe tomography showed that high number densities of He bubbles were formed on the surface of nanoclusters and Ti(C,N) precipitates, and along grain boundaries and dislocations. At higher fluences, facetted bubbles are formed and it is postulated that the lowest energy state configuration is the truncated rhombic dodecahedron.

  17. Tensile response of low activation ferritic steels irradiated in ORR at temperatures in the range 60-400 °C

    NASA Astrophysics Data System (ADS)

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

    2002-12-01

    Post-irradiation tensile test results are reported for a series of low activation steels containing manganese following irradiation in the Oak Ridge Reactor at 60, 200, 330 and 400 °C to ˜10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5% and W to 1.0%. Strengths are higher in all alloys for irradiation conditions below 400 °C, with peak hardening occurring following irradiation at 200 °C. The 9Cr alloy class exhibited the smallest increases in hardening. Test results were consistent with previous results obtained on fast flux test facility-irradiated specimens. Manganese does not appear to play a role in the hardening observed at these low irradiation temperatures.

  18. 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.

  19. Applicability of copper alloys for DEMO high heat flux components

    NASA Astrophysics Data System (ADS)

    Zinkle, Steven J.

    2016-02-01

    The current state of knowledge of the mechanical and thermal properties of high-strength, high conductivity Cu alloys relevant for fusion energy high heat flux applications is reviewed, including effects of thermomechanical and joining processes and neutron irradiation on precipitation- or dispersion-strengthened CuCrZr, Cu-Al2O3, CuNiBe, CuNiSiCr and CuCrNb (GRCop-84). The prospects for designing improved versions of wrought copper alloys and for utilizing advanced fabrication processes such as additive manufacturing based on electron beam and laser consolidation methods are discussed. The importance of developing improved structural materials design criteria is also noted.

  20. 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.