Irradiation Creep

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Gary S. Was - One of the best experts on this subject based on the ideXlab platform.

  • Irradiation Creep and growth
    2017
    Co-Authors: Gary S. Was
    Abstract:

    Creep is the time-dependent deformation of a metal under constant load and at high temperature (T/T m > 0.3). The metal responds by elongating with a strain defined as either the nominal strain, e, calculated from the original length of the sample.

  • proton Irradiation Creep of fm steel t91
    Journal of Nuclear Materials, 2015
    Co-Authors: Gary S. Was
    Abstract:

    Abstract Ferritic–martensitic (FM) steel T91 was subjected to Irradiation with 3 MeV protons while under load at stresses of 100–200 MPa, temperatures between 400 °C and 500 °C, and dose rates between 1.4 × 10 −6  dpa/s and 5 × 10 −6  dpa/s to a total dose of less than 1 dpa. Creep behavior was analyzed for parametric dependencies. The temperature dependence was found to be negligible between 400 °C and 500 °C, and the dose rate dependence was observed to be linear. Creep rate was proportional to stress at low stress values and varied with stress to the power 14 above 160 MPa. The large stress exponent of the proton Irradiation Creep experiments under high stress suggested that dislocation glide was driving both thermal and Irradiation Creep. Microstructure observations of anisotropic dislocation loops also contributed to the total Creep strain. After subtracting the power law Creep and anisotropic dislocation loop contributions, the remaining Creep strain was accounted for by dislocation climb enabled by stress induced preferential absorption (SIPA) and preferential dislocation glide (PAG).

  • proton Irradiation Creep of beta silicon carbide
    Journal of Nuclear Materials, 2011
    Co-Authors: Vani Shankar, Gary S. Was
    Abstract:

    Abstract In situ Irradiation Creep behavior of chemically vapor-deposited (CVD) polycrystalline beta silicon carbide (β-SiC) has been studied using proton beam of energies 2.8 MeV and 3.2 MeV. Experiments were conducted at 1183 K and at stresses of 18.5 MPa and 97.9 MPa between dose rates of 1.5 and 2.45 × 10−6 dpa/s. Strain was measured using a laser speckle extensometer (LSE) and a linear variable differential transformer (LVDT), and temperature was measured using a 2-dimensional infrared pyrometer. Results showed that the total strain rate increased with increasing stress and dose rate. Shifts of XRD peaks following proton Irradiation of SiC at 1183 K indicated that swelling had occurred and that it increased with dose. A uniform expansion of the lattice with no X-ray line broadening clearly indicated that the swelling at doses up to 0.37 dpa was due to single point defects. The swelling rate was determined and subtracted from the measured total strain rate to obtain the true Creep rate. The Creep rate was found to exhibit a linear dependence on the applied tensile stress, and on dose rate to the third power.

  • in situ proton Irradiation Creep of ferritic martensitic steel t91
    Transactions of the American Nuclear Society, 2010
    Co-Authors: Gary S. Was
    Abstract:

    Abstract An Irradiation Creep apparatus was developed for in situ straining of T91 strip samples while exposed to 2–3 MeV proton Irradiation at 300–600 °C. Thermal Creep experiments were conducted at 600 °C, 47 MPa, and 500 °C, 160 MPa. The thermal Creep strains were in reasonable agreement with literature data on bulk samples of T91. An Irradiation Creep experiment was conducted at 500 °C and 160 MPa with a damage rate range from 3.1 × 10 −6  dpa/s to 4.9 × 10 −6  dpa/s. The Creep rate of T91 was found to increase linearly with dose rate. A TEM investigation of the irradiated microstructure showed signs of dislocation pileup, subgrain formation, and small dislocation loops. The results illustrate the utility of accelerator-Creep experiments to obtain Creep rates at low dose and the capability to observe transient changes in real time, thus providing the tools for isolating the effects of individual variables on Creep rate of T91.

  • fundamentals of radiation materials science metals and alloys
    2007
    Co-Authors: Gary S. Was
    Abstract:

    Radiation Damage.- The Radiation Damage Event.- The Displacement of Atoms.- The Damage Cascade.- Point Defect Formation and Diffusion.- Radiation-Enhanced Diffusion and Defect Reaction Rate Theory.- Physical Effects of Radiation Damage.- Radiation-Induced Segregation.- Dislocation Microstructure.- Irradiation-Induced Voids and Bubbles.- Phase Stability Under Irradiation.- Unique Effects of Ion Irradiation.- Simulation of Neutron Irradiation Effects with Ions.- Mechanical Effects of Radiation Damage.- Irradiation Hardening and Deformation.- Fracture and Embrittlement.- Irradiation Creep and Growth.- Environmentally Assisted Cracking of Irradiated Metals and Alloys.

F.a. Garner - One of the best experts on this subject based on the ideXlab platform.

  • Irradiation Creep and density changes observed in ma957 pressurized tubes irradiated to doses of 40 110 dpa at 400 750 c in fftf
    Journal of Nuclear Materials, 2012
    Co-Authors: Mychailo B Toloczko, F.a. Garner, S A Maloy
    Abstract:

    Abstract An Irradiation Creep and swelling study was performed on tubing constructed from the yttrium/titanium oxide dispersion strengthened (ODS) ferritic steel MA957. As a result of the reduction operations during manufacture, the grains in the tubing were highly elongated in the direction of the tubing longitudinal axis. Pressurized Creep tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 dpa to 110 dpa at target temperatures ranging from 400 to 750 °C. The diametral strains produced during Irradiation exhibit primary (transient) Creep strains that are dependent on stress and increase with Irradiation temperature and are followed by a temperature-independent steady-state Creep rate of ∼0.75 × 10 −6  (MPa dpa) −1 , a value similar to that of traditional tempered ferritic/martensitic steels. Contributions to primary Creep strains may arise not only from classical thermal Creep or Irradiation Creep considerations, but also may result from an Irradiation-stimulated growth process whereby the highly elongated grain structure shrinks somewhat in the elongated direction, reducing the tubing aspect ratio to produce slightly fatter grains and thereby increasing the tube diameter. One manifestation of this process is a change in tube diameter that is not accompanied by a density change characteristic of either void swelling or precipitation-induced changes in lattice parameter. These results provide the first demonstration that resistance to Irradiation Creep can be extended to higher temperatures by dispersoid addition, and most importantly, this resistance is maintained to high radiation damage levels at least for temperatures of 600 °C or less.

  • swelling and Creep observed in aisi 304 fuel pin cladding from three mox fuel assemblies irradiated in ebr ii
    Journal of Nuclear Materials, 2011
    Co-Authors: F.a. Garner, B.j. Makenas, S A Chastain
    Abstract:

    Abstract Three 37-pin MOX-fueled experimental subassemblies were irradiated in EBR-II with fuel pin cladding constructed from annealed AISI 304 stainless steel. Analysis of the swelling and Irradiation Creep of the cladding showed that the terminal swelling rate of AISI 304 stainless steel appears to be ∼1%/dpa and that swelling is very reproducible for identical Irradiation conditions. The swelling at a given neutron fluence is rather sensitive to both Irradiation temperature and especially to the neutron flux, however, with the primary influence residing in the transient regime. As the neutron flux increases the duration of the transient regime is increased in agreement with other recent studies. The duration of the transient regime is also decreased by increasing Irradiation temperature. In these assemblies swelling reached high levels rather quickly, reducing the opportunity for fuel pin cladding interaction and thereby reducing the contribution of Irradiation Creep to the total deformation. It also appears that in this swelling-before-Creep scenario that the well-known “Creep disappearance” phenomenon was operating strongly.

  • 10 void swelling and Irradiation Creep in light water reactor lwr environments
    Understanding and Mitigating Ageing in Nuclear Power Plants#R##N#Materials and Operational Aspects of Plant Life Management (Plim), 2010
    Co-Authors: F.a. Garner
    Abstract:

    : Until 1993 it was assumed that swelling and Irradiation Creep were phenomena of little importance to light water cooled reactors. It is now recognized that swelling and Irradiation Creep are in progress in austenitic internals of pressurized water reactors (PWRs) especially, with boiling water reactors (BWRs) not being as vulnerable to these processes. Some manifestations of swelling and Irradiation Creep are already being observed in PWRs. Owing to the non-linear development of swelling with increasing neutron exposure, it is expected that consequences of swelling and Irradiation Creep will accelerate, especially as PWRs move beyond their original design lives of forty years.

  • the influence of cold work level on the Irradiation Creep and swelling of aisi 316 stainless steel irradiated as pressurized tubes in the ebr ii fast reactor
    Journal of Nuclear Materials, 2007
    Co-Authors: Edgar R Gilbert, F.a. Garner
    Abstract:

    Pressurized tubes of AISI 316 stainless steel irradiated in the P-1 experiment in the EBR-II fast reactor have been measured to determine the dependence of Irradiation-induced strains resulting from plastic deformation, Irradiation Creep, void swelling and precipitation. It is shown that the Soderberg relation predicting no axial Creep strains in biaxially-loaded tubes is correct for both plastic and Creep strains. Swelling strains are shown to be isotropically distributed both for stress-free and stress-affected swelling, while precipitation strains are somewhat anisotropic in their distribution. When corrected for stress-enhancement of swelling, the derived Irradiation Creep strains appear to be identical for both annealed and 20% cold-worked specimens, and also for tubes strained by rise to power increases in pressure. For relatively small Creep strains it is often difficult to separate the Creep and non-Creep components of deformation.

  • swelling Irradiation Creep and growth of pure rhenium irradiated with fast neutrons at 1030 1330 c
    Journal of Nuclear Materials, 2000
    Co-Authors: F.a. Garner, Mychailo B Toloczko, Lawrence R Greenwood, Cheryl R Eiholzer, M M Paxton, Raymond J Puigh
    Abstract:

    Abstract This paper discusses the results of two series of experiments conducted on pure hcp rhenium in the EBR-II and FFTF fast reactors. In FFTF, density change data were derived from open tubes and solid rods irradiated at temperatures and fluences in the range of 1020–1250°C and 4.4–8.3×1022 n cm−2, respectively (E > 0.1 MeV). Both density change and diametral change data were obtained from pressurized tubes irradiated in EBR-II to ∼0.65 and ∼5.1×1022 n cm−2 at temperatures between 1030°C and 1330°C. Analysis of the data shows that four concurrent processes contribute to the radiation-induced strains observed in these experiments. These are void swelling, transmutation-induced densification via production of osmium, Irradiation Creep and Irradiation growth.

Wolfgang Hoffelner - One of the best experts on this subject based on the ideXlab platform.

  • Irradiation Creep of candidate materials for advanced nuclear plants
    Journal of Nuclear Materials, 2013
    Co-Authors: Jiachao Chen, P Jung, Wolfgang Hoffelner
    Abstract:

    Abstract In the present paper, Irradiation Creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ Irradiation Creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2–8 × 10 −6  dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the Irradiation Creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to Irradiation Creep models.

  • Irradiation Creep and microstructural changes of ods steels of different cr contents during helium implantation under stress
    Journal of Nuclear Materials, 2013
    Co-Authors: Jiachao Chen, P Jung, J Henry, Y De Carlan, T Sauvage, F Duval, M F Barthe, Wolfgang Hoffelner
    Abstract:

    Abstract Irradiation Creep and microstructural changes of two ferritic ODS steels with 12% and 14% Cr have been studied by homogeneously implantation with helium under uniaxial tensile stresses from 40 to 300 MPa. The maximum dose was about 1.2 dpa (5000 appm-He) with displacement damage rates of 1 × 10 −5  dpa/s at a temperature of 300 °C. Irradiation Creep compliances were measured to be 4.0 × 10 −6  dpa −1  MPa −1 and 10 × 10 −6  dpa −1  MPa −1 for 12 and 14Cr ODS, respectively. Subsequently, microstructural evolution was studied in detail by TEM observations, showing dislocation loops and bubbles distributed homogenously in the matrix. Some bubbles were attached to ODS particles. Finally, the effects of Cr content on Irradiation Creep and microstructural changes are discussed, including earlier results of a 19Cr ODS and a PM2000 ferritic steel.

  • Damage assessment in structural metallic materials for advanced nuclear plants
    Journal of Materials Science, 2010
    Co-Authors: Wolfgang Hoffelner
    Abstract:

    Future advanced nuclear plants are considered to operate as cogeneration plants for electricity and heat. Metals and alloys will be the main portion of structural materials employed (including fuel claddings). Due to the operating conditions these materials are exposed to damaging conditions like Creep, fatigue, Irradiation and its combinations. The paper uses the most important alloys: ferritic-martensitic steels, superalloys, oxide dispersion strengthened steels and to some extent titanium aluminides to discuss its responses to these exposure conditions. Extrapolation of stress rupture data, Creep strain, swelling, Irradiation Creep and Creep–fatigue interactions are considered. Although the stress rupture- and the Creep behavior seem to meet expectations, the long design lives of 60 years are really challenging for extrapolations and particularly questions like negligible Creep or occurrence of diffusion Creep need special attention. Ferritic matrices (including oxide dispersion strengthened (ODS), steels) have better Irradiation swelling behavior than austenites. Presence and size of dispersoids having a strong influence on high-temperature strength bring only insignificant improvements in Irradiation Creep. A strain-range-separation based approach for Creep–fatigue interactions is presented which allows a real prediction of Creep–fatigue lives. An assessment of capabilities and limitations of advanced materials modeling tools with respect to damage development is given.

  • Irradiation Creep of oxide dispersion strengthened ods steels for advanced nuclear applications
    Journal of Nuclear Materials, 2009
    Co-Authors: Jiachao Chen, Wolfgang Hoffelner
    Abstract:

    Abstract Ferritic oxide dispersion strengthened steels with different microstructure were in-beam Creep tested in a temperature range from 300 to 500 °C. Irradiation was by He-ions. Elongation was determined as a function of stress and Irradiation damage rate. Damage was investigated by transmission electron microscopy. A thorough analysis of the loops developing during Irradiation Creep did not show any dependence of orientation or size on the direction of the applied stress. At 400 °C radiation induced segregation was found (most probably an iron aluminide) which had no effect on Irradiation Creep. No pronounced influence of microstructure or dispersoid size on the Irradiation Creep behavior was detected. Irradiation Creep compliance of PM2000 with dispersoids of about 30 nm diameter were found to differ little from material with dispersoids of only 2–3 nm diameter. This is in contrast to thermal Creep where dislocation–obstacle interactions are extremely important. An assessment of the technical relevance of Irradiation Creep in advanced nuclear systems is presented.

  • Irradiation Creep and microstructural changes in an advanced ods ferritic steel during helium implantation under stress
    Journal of Nuclear Materials, 2009
    Co-Authors: Jiachao Chen, P Jung, Manuel A Pouchon, Akihiko Kimura, Wolfgang Hoffelner
    Abstract:

    An advanced oxide dispersion strengthened (ODS) ferritic steel with very fine oxide particles has been homogeneously implanted with helium under uniaxial tensile stresses from 20 to 250 MPa to a maximum dose of about 0.38 dpa (1650 appm-He) with displacement damage rates of 4.4 × 10−6 dpa/s at temperatures of 573 and 773 K. The samples were in the form of miniaturized dog-bones, where during the helium implantation the straining and the electrical resistance were monitored simultaneously. Creep compliances were measured to be 4.0 × 10−6 and 11 × 10−6 dpa−1 MPa−1 at 573 and 773 K, respectively. The resistivity of ODS steel samples decreased with dose, indicating segregation and/or precipitation. Evolution of microstructure during helium implantation was studied in detail by TEM. The effects of ODS particle size on Irradiation Creep and microstructural changes was investigated by comparing the results from the present advanced ODS (K1) to a commercial ODS ferritic steels (PM2000) with much bigger oxide particles.

F. Onimus - One of the best experts on this subject based on the ideXlab platform.

Bréchet Yves - One of the best experts on this subject based on the ideXlab platform.

  • In-situ TEM Irradiation Creep experiment revealing radiation induced dislocation glide in pure copper
    'Elsevier BV', 2021
    Co-Authors: Khiara Nargisse, Onimus Fabien, Jublot-leclerc Stéphanie, Jourdan Thomas, Pardoen Thomas, Raskin Jean-pierre, Bréchet Yves
    Abstract:

    In-situ TEM straining experiments were performed on pure copper to investigate the dislocation activity under heavy ion Irradiation and high applied stress levels. The unpinning of dislocations from Irradiation defects followed by glide was observed under Irradiation at stress levels just below the critical stress for dislocation glide without Irradiation. This phenomenon, unraveled for the first time in copper, has been statistically analyzed using digital image processing. Quantitative analysis of pinning lifetimes has been performed, suggesting that a cascade related mechanism is operative to explain the radiation induced dislocation glide. This work provides new insights on the Irradiation Creep deformation at high stress level

  • In-situ TEM Irradiation Creep experiment revealing radiation induced dislocation glide in pure copper
    'Elsevier BV', 2021
    Co-Authors: Khiara Nargisse, Onimus Fabien, Jublot-leclerc Stéphanie, Jourdan Thomas, Pardoen Thomas, Raskin Jean-pierre, Bréchet Yves
    Abstract:

    International audienceIn-situ straining experiments were performed on pure copper to investigate dislocation motion under heavy ion Irradiation at high stress levels. The unpinning of dislocations from Irradiation defects followed by glide was observed under Irradiation at stress level just below the critical stress for dislocation glide without Irradiation. This phenomenon was unraveled for the first time in copper. The dislocation dynamics recorded in-situ was statistically analyzed using digital image processing to determine the pinning lifetime. Quantitative analysis of pinning lifetimes have been performed, suggesting that a cascade related mechanism is operative to explain the fast dislocation glide under Irradiation. This work provides a new insight on the Irradiation Creep deformation at high stress level

  • A novel displacement cascade driven Irradiation Creep mechanism in α-zirconium: A molecular dynamics study
    'Elsevier BV', 2020
    Co-Authors: Khiara Nargisse, Onimus Fabien, Pardoen Thomas, Raskin Jean-pierre, Dupuy Laurent, Kassem Wassim, Crocombette Jean-paul, Bréchet Yves
    Abstract:

    Zirconium alloys used in nuclear reactors undergo Irradiation Creep, which consists of a visco-plastic deformation activated by Irradiation occurring under constant load. However, the fundamental underlying mechanisms have not been unraveled yet. A new high-stress Irradiation Creep mechanism for recrystallized Zircaloy-4 has recently been proposed based on in situ ion Irradiation deformation experiments. A displacement cascade is assumed to induce a direct unpinning of a dislocation from an Irradiation defect if the cascade occurs within an effective volume around the pinning point. In the present work, a systematic molecular dynamics study was performed to investigate the effect of a 20 keV cascade occurring near -screw dislocation pinned on an interstitial -loop. The direct release of dislocations by displacement cascades is predicted by the simulations. This release is more likely around the pinning points and with increasing stress, in agreement with experimental observations. The effective volume is roughly estimated for three stress levels equal to 0.89tc, 0.93tc and 0.97tc, with tc