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Sample records for neutron-irradiated cr-mo ferritic

  1. 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.

  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. 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.

  4. Type IV Cracking Susceptibility in Weld Joints of Different Grades of Cr-Mo Ferritic Steel

    NASA Astrophysics Data System (ADS)

    Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Bhanu Sankara Rao, K.

    2009-02-01

    Relative type-IV cracking susceptibility in 2.25Cr-1Mo, 9Cr-1Mo, and 9Cr-1MoVNb ferritic steel weld joint has been assessed. The type-IV cracking was manifested as preferential accumulation of creep deformation and cavitation in the relatively soft intercritical region of heat affected zone of the weld joint. The type-IV cracking susceptibility has been defined as the reduction in creep-rupture strength of weld joint compared to its base metal. The 2.25Cr-1Mo steel exhibited more susceptibility to type-IV cracking at relatively lower temperatures; whereas, at higher temperatures, 9Cr-1MoVNb steel was more susceptible. The relative susceptibility to type-IV cracking in the weld joint of the Cr-Mo steels has been rationalized on the basis of creep-strengthening mechanisms operating in the steels and their venerability to change on intercritical heating during weld thermal cycle, subsequent postweld heat treatment, and creep exposure.

  5. Effect of Niobium on the Ferrite Continuous-Cooling-Transformation (CCT) Curve of Ultrahigh-Thickness Cr-Mo Steel

    NASA Astrophysics Data System (ADS)

    Lee, Sanghoon; Na, Hyesung; Kim, Byunghoon; Kim, Dongjin; Kang, Chungyun

    2013-06-01

    Pressure vessels made for petrochemical and power plants using Cr-Mo steel must be thick (≥400 mm) and have high tensile strength (≥600 MPa). However, the tensile strength in the middle portion of the vessel is very low as a result of ferrite formation. Therefore, research was performed to study the ferrite transformation that occurs in the middle portion of high-thickness Cr-Mo steel when Nb is added to it. The ferrite-formation start time of the continuous-cooling-transformation (CCT) curve decreased with an increase in Nb content until the latter reached 0.05 pct. On cooling from the austenitizing temperature, some of the NbC present at the austenitizing temperature of 1203 K (930 °C) goes into austenite solution in the temperature range of 1173 K to 1073 K (900 °C to 800 °C). However, the ferrite curve shifted to the left for the alloy containing 0.075 pct Nb. It is postulated that the surplus NbC could act as ferrite nucleation sites despite the lower cooling rate. As a result, the hardenability improved in the order of the following Nb content: 0.05 pct, 0.025 pct, 0 pct, and 0.075 pct.

  6. Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1997-08-01

    Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.

  7. Analysis of Creep Rupture Behavior of Cr-Mo Ferritic Steels in the Presence of Notch

    NASA Astrophysics Data System (ADS)

    Goyal, Sunil; Laha, K.; Das, C. R.; Mathew, M. D.

    2015-01-01

    Effect of notch on creep rupture behavior of 2.25Cr-1Mo, 9Cr-1Mo, and modified 9Cr-1Mo ferritic steels has been assessed. Creep tests were carried out on smooth and notched specimens of the steels in the stress ranging 90 to 300 MPa at 873 K (600 °C). Creep rupture lives of the steels increased in the presence of notch over those of smooth specimens, thus exhibiting notch strengthening. The strengthening was comparable for the 9Cr-1Mo and 2.25Cr-1Mo steels and appreciably more in modified 9Cr-1Mo steel. The strengthening effect was found to decrease with the decrease in applied stress and increase in rupture life for all the steels. The presence of notch decreased the creep rupture ductility of the steels significantly and the 2.25Cr-1Mo steel suffered more reduction than in the other two 9Cr-steels. Finite element analysis of stress distribution across the notch was carried out to understand the notch strengthening and its variation in the steels. The variation in fracture appearance has also been corroborated based on finite element analysis. Reduction in von-Mises stress across the notch throat plane resulted in strengthening in the steels. Higher reduction in von-Mises stress in modified 9Cr-1Mo steel than that in 2.25Cr-1Mo and 9Cr-1Mo steels induced more strengthening in modified 9Cr-1Mo steel under multiaxial state of stress.

  8. Effects of neutron irradiation on microstructural evolution in candidate low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kohno, Yutaka; Kohyama, Akira; Yoshino, Masahiko; Asakura, Kentaro

    1994-09-01

    Fe-(2.25-12)Cr-2W-V, Ta low activation ferritic steels (JLF series steels) were developed in the fusion materials development program of Japanese universities. Microstructural observations, including precipitation response, were performed after neutron irradiation in the FFTF/MOTA. The preirradiation microstructure was stable after irradiation at low temperature (< 683 K). Recovery of martensitic lath structure and coarsening of precipitates took place above 733 K. Precipitates observed after irradiation were the same as those in unirradiated materials in 7-9Cr steels, and no irradiation induced phase was identified. The irradiation induced shift in DBTT in the 9Cr-2W steel proved to be very small which is a reflection of stable precipitation response in these steels. A high density of fine α' precipitates was observed in the 12Cr steel which might be responsible for the large irradiation hardening found in the 12Cr steel. Void formation was observed in 7-9Cr steels irradiated at 683 K, but the amount of void swelling was very small.

  9. Mechanical property changes of low activation ferritic/martensitic steels after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M. L.; Narui, M.

    Mechanical property changes of Fe- XCr-2W-0.2V,Ta ( X: 2.25-12) low activation ferritic/martensitic steels including Japanese Low Activation Ferritic/martensitic (JLF) steels and F82H after neutron irradiation were investigated with emphasis on Charpy impact property, tensile property and irradiation creep properties. Dose dependence of ductile-to-brittle transition temperature (DBTT) in JLF-1 (9Cr steel) irradiated at 646-700 K increased with irradiation up to 20 dpa and then decreased with further irradiation showing highest DBTT of 260 K at 20 dpa. F82H showed similar dose dependence in DBTT to JLF-1 with higher transition temperature than that of JLF-1 at the same displacement damage. Yield strength in JLF steels and F82H showed similar dose dependence to that of DBTT. Yield strength increased with irradiation up to 15-20 dpa and then decreased to saturate above about 40 dpa. Irradiation hardening in 7-9%Cr steels (JLF-1, JLF-3, F82H) were observed to be smaller than those in steels with 2.25%Cr (JLF-4) or 12%Cr (JLF-5). Dependences of creep strain on applied hoop stress and neutron fluence were measured to be 1.5 and 1, respectively. Temperature dependence of creep coefficient showed a maximum at about 700 K which was caused by irradiation induced void formation or irradiation enhanced creep deformation. Creep coefficient of F82H was larger than those of JLF steels above 750 K. This was considered to be caused by the differences in N and Ta concentration between F82H and JLF steels.

  10. Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

    SciTech Connect

    Ahmad Alsabbagh; Apu Sarkar; Brandon Miller; Jatuporn Burns; Leah Squires; Douglas Porter; James I. Cole; K. L. Murty

    2014-10-01

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

  11. Low cycle fatigue properties of reduced activation ferritic/martensitic steels after high-dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Aktaa, J.; Povstyanko, A.; Prokhorov, V.; Diegele, E.; Lässer, R.

    2011-08-01

    This paper focuses on the low cycle fatigue (LCF) behaviour of reduced activation ferritic/martensitic steels irradiated to a displacement damage dose of up to 70 dpa at 330-337 °C in the BOR 60 reactor within the ARBOR 2 irradiation programme. The influence of neutron irradiation on the fatigue behaviour was determined for the as-received EUROFER97, pre-irradiation heat-treated EUROFER97 HT and F82H-mod steels. Strain-controlled push-pull loading was performed using miniaturized cylindrical specimens at a constant temperature of 330 °C with total strain ranges between 0.8% and 1.1%. Comparison of the LCF behaviour of irradiated and reference unirradiated specimens was performed for both the adequate total and inelastic strains. Neutron irradiation-induced hardening may have various effects on the fatigue behaviour of the steels. The reduction of inelastic strain in the irradiated state compared with the reference unirradiated state at common total strain amplitudes may increase fatigue lifetime. The increase in the stress at the adequate inelastic strain, by contrast, may accelerate fatigue damage accumulation. Depending on which of the two effects mentioned dominates, neutron irradiation may either extend or reduce the fatigue lifetime compared with the reference unirradiated state. The results obtained for EUROFER97 and EUROFER97 HT confirm these considerations. Most of the irradiated specimens show fatigue lifetimes comparable to those of the reference unirradiated state at adequate inelastic strains. Some irradiated specimens, however, show lifetime reduction or increase in comparison with the reference state at adequate inelastic strains.

  12. Effect of neutron irradiation at low temperature on the embrittlement of the reduced-activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Rybin, V. V.; Kursevich, I. P.; Lapin, A. N.

    1998-10-01

    Effects of neutron irradiation to fluence of 2.0 × 10 24 n/m 2 ( E > 0.5 MeV) in temperature range 70-300°C on mechanical properties and structure of the experimental reduced-activation ferritic 0.1%C-(2.5-12)%Cr-(1-2)%W-(0.2-0.7)%V alloys were investigated. The steels were studied in different initial structural conditions obtained by changing the modes of heat treatments. Effect of neutron irradiation estimated by a shift in ductile-brittle transition temperature (ΔDBTT) and reduction of upper shelf energy (ΔUSE) highly depends on both irradiation condition and steel chemical composition and structure. For the steel with optimum chemical composition (9Cr-1.5WV) after irradiation to 2 × 10 24 n/m 2 ( E ⩾ 0.5 MeV) at 280°C the ΔDBTT does not exceed 25°C. The shift in DBTT increased from 35°C to 110°C for the 8Cr-1.5WV steel at a decrease in irradiation temperature from 300°C to 70°C. The CCT diagrams are presented for several reduced-activated steels.

  13. 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.

  14. A Comparison of Creep Rupture Strength of Ferritic/Austenitic Dissimilar Weld Joints of Different Grades of Cr-Mo Ferritic Steels

    NASA Astrophysics Data System (ADS)

    Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Goyal, Sunil; Mathew, M. D.

    2012-04-01

    Evaluations of creep rupture properties of dissimilar weld joints of 2.25Cr-1Mo, 9Cr-1Mo, and 9Cr-1MoVNb steels with Alloy 800 at 823 K were carried out. The joints were fabricated by a fusion welding process employing an INCONEL 182 weld electrode. All the joints displayed lower creep rupture strength than their respective ferritic steel base metals, and the strength reduction was greater in the 2.25Cr-1Mo steel joint and less in the 9Cr-1Mo steel joint. Failure location in the joints was found to shift from the ferritic steel base metal to the intercritical region of the heat-affected zone (HAZ) of the ferritic steel (type IV cracking) with the decrease in stress. At still lower stresses, the failure in the joints occurred at the ferritic/austenitic weld interface. The stress-life variation of the joints showed two-slope behavior and the slope change coincided with the occurrence of ferritic/austenitic weld interface cracking. Preferential creep cavitation in the soft intercritical HAZ induced type IV failure, whereas creep cavitation at the interfacial particles induced ferritic/austenitic weld interface cracking. Micromechanisms of the type IV failure and the ferritic/austenitic interface cracking in the dissimilar weld joint of the ferritic steels and relative cracking susceptibility of the joints are discussed based on microstructural investigation, mechanical testing, and finite element analysis (FEA) of the stress state across the joint.

  15. 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.

  16. 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.

  17. Relationship Between Grain Boundary Structure and Radiation Induced Segregation in a Neutron Irradiated 9 wt. % Cr Model Ferritic/Martensitic Steel

    SciTech Connect

    Field, Kevin G; Miller, Brandon; Chichester, Heather J.M.; Sridharan, K.; Allen, Todd R.

    2014-01-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs but has only been demonstrated in ion irradiated specimens. A 9 wt. % Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of neutron radiation environment on the RIS-GB structure dependence. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  18. Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9 wt.% Cr model ferritic/martensitic steel

    SciTech Connect

    Field, Kevin G.; Miller, Brandon D.; Chichester, Heather J. M.; Sridharan, Kumar; Allen, Todd R.

    2014-02-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  19. A high Cr-Mo alloy iron

    SciTech Connect

    Huang, X.; Wu, Y.

    1998-08-01

    The structure, hardness, abrasion, and erosion wear of Cr-Mo white iron (containing approximately 28% Cr and 1% Mo) heat treated at certain temperatures were studied. Results show that the heat treatment of white iron changes the structures and properties; that is morphology, amount, size, and distribution of secondary phases are affected. When white iron was heated at 800 to 850 C the secondary phase precipitated at the phase boundary, making the abrasion and erosion wear worse. When the iron was heated at 900 to 950 C, the secondary phase precipitated dispersively at the matrix, and the corrosion wear was optimum. If the iron is heated at 1000 to 1050 C, the resistance of abrasion is inhibited, as the secondary phases precipitate in large amounts, and the hardness is increased. When the white iron is tempered at 500 to 600 C, the resistance of abrasion is better.

  20. 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.

  1. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  2. Effect of nickel on the precipitation processes in 12CrMoV steels under creep at 550 C

    SciTech Connect

    Vodarek, V.; Strang, A.

    1997-12-18

    In the course of development of 12 CrMoV steels nickel was added in order to improve impact properties and to suppress the presence of {delta}-ferrite in the microstructure. It was found, however, that excessive amounts of nickel, greater than {approximately}0.6 wt. %, caused an accelerated reduction in the creep rupture strength. This resulted in a downward inflexion in time dependence of creep rupture strength which is known as sigmoidal behavior. This phenomenon was particularly evident when these steels were tested at temperatures of 550 C and greater. In order to improve understanding of the effect of nickel on the microstructural evolution of 12CrMoV steels the metallographic investigation was undertaken on a series of four casts with different contents of nickel, which had been creep tested at 550 C out to durations of around 100,000 hours.

  3. Microscopic impact of creep damage incipience and development on the magnetic properties of ferromagnetic Cr Mo steel

    NASA Astrophysics Data System (ADS)

    Augustyniak, Boleslaw; Piotrowski, Leszek; Chmielewski, Marek; Sablik, Martin J.

    2006-09-01

    Results are presented for magnetoacoustic emission (MAE) and magnetostriction measurements for Cr-Mo P22 steel (2.25Cr-1Mo) in the as-manufactured stage and after service at a power plant. We argue that MAE is correlated with magnetostriction behavior in relation to its derivative. Some quantitative description of precipitate morphology elucidates the as-observed decrease of MAE intensity with creep damage by change of precipitate morphology in the ferrite grain and grain boundary. This is discussed via a simple model for the MAE intensity dependence on microstructural change.

  4. Activity of enzyme immobilized on silanized Co-Cr-Mo.

    PubMed

    Puleo, D A

    1995-08-01

    The surface of an orthopedic biomaterial was modified by the covalent immobilization of biomolecules. Derivatization of Co-Cr-Mo samples with organic and aqueous solutions of gamma-aminopropyltriethoxysilane (APS) resulted in a concentration-dependent number of reactive NH2 groups on the surface available for coupling to protein. The enzyme trypsin was used as a model biomolecule to investigate the effect of immobilization on proteolytic activity. Trypsin was coupled to the silanized samples by formation of Schiff's base linkages via glutaraldehyde. The nature of the interaction between trypsin and biomaterial was then probed by treatment with concentrated guanidine hydrochloride (GuHCl) and urea. Residual activity (following treatment with chaotropic agents) of trypsin immobilized on silanized Co-Cr-Mo was dependent both on the nature of the silane solution and on the type of chaotropic agent. Organic silanization with APS required a minimum density of approximately 49 NH2 per nm2 of nominal surface area (> 0.021 M APS) for residual activity of immobilized trypsin. For aqueous silanization, approximately 5.4 NH2/nm2 (0.51 M APS) resulted in maximal residual trypsin activity. Treatment with GuHCl removed more trypsin activity from Co-Cr-Mo samples silanized with organic solutions of APS than did treatment with urea. On the contrary, with aqueous silanization the samples possessed greater residual activity following treatment with GuHCl than following urea. Compared to simple adsorption with protein onto Co-Cr-Mo, both methods of silanization with APS resulted in superior residual immobilized enzyme activity. PMID:7593038

  5. Albumin adsorption on CoCrMo alloy surfaces

    PubMed Central

    Yan, Yu; Yang, Hongjuan; Su, Yanjing; Qiao, Lijie

    2015-01-01

    Proteins can adsorb on the surface of artificial joints immediately after being implanted. Although research studying protein adsorption on medical material surfaces has been carried out, the mechanism of the proteins’ adsorption which affects the corrosion behaviour of such materials still lacks in situ observation at the micro level. The adsorption of bovine serum albumin (BSA) on CoCrMo alloy surfaces was studied in situ by AFM and SKPFM as a function of pH and the charge of CoCrMo alloy surfaces. Results showed that when the specimens were uncharged, hydrophobic interaction could govern the process of the adsorption rather than electrostatic interaction, and BSA molecules tended to adsorb on the surfaces forming a monolayer in the side-on model. Results also showed that adsorbed BSA molecules could promote the corrosion process for CoCrMo alloys. When the surface was positively charged, the electrostatic interaction played a leading role in the adsorption process. The maximum adsorption occurred at the isoelectric point (pH 4.7) of BSA. PMID:26673525

  6. Albumin adsorption on CoCrMo alloy surfaces

    NASA Astrophysics Data System (ADS)

    Yan, Yu; Yang, Hongjuan; Su, Yanjing; Qiao, Lijie

    2015-12-01

    Proteins can adsorb on the surface of artificial joints immediately after being implanted. Although research studying protein adsorption on medical material surfaces has been carried out, the mechanism of the proteins’ adsorption which affects the corrosion behaviour of such materials still lacks in situ observation at the micro level. The adsorption of bovine serum albumin (BSA) on CoCrMo alloy surfaces was studied in situ by AFM and SKPFM as a function of pH and the charge of CoCrMo alloy surfaces. Results showed that when the specimens were uncharged, hydrophobic interaction could govern the process of the adsorption rather than electrostatic interaction, and BSA molecules tended to adsorb on the surfaces forming a monolayer in the side-on model. Results also showed that adsorbed BSA molecules could promote the corrosion process for CoCrMo alloys. When the surface was positively charged, the electrostatic interaction played a leading role in the adsorption process. The maximum adsorption occurred at the isoelectric point (pH 4.7) of BSA.

  7. Quantitative TEM analysis of precipitation and grain boundary segregation in neutron irradiated EUROFER 97

    NASA Astrophysics Data System (ADS)

    Dethloff, Christian; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-01

    Characterization of irradiation induced microstructural defects is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM) is used to analyze the types and structure of precipitates, and the evolution of their size distributions and densities caused by neutron irradiation to a dose of 32 displacements per atom (dpa) at 330-340 °C in the irradiation experiment ARBOR 1. A significant growth of MX and M23C6 type precipitates is observed after neutron irradiation, while the precipitate density remains unchanged. Hardening caused by MX and M23C6 precipitate growth is assessed by applying the Dispersed Barrier Hardening (DBH) model, and shown to be of minor importance when compared to other irradiation effects like dislocation loop formation. Additionally, grain boundary segregation of chromium induced by neutron irradiation was investigated and detected in irradiated specimens.

  8. CoCrMo Metal-on-Metal Hip Replacements

    PubMed Central

    Liao, Yifeng; Hoffman, Emily; Wimmer, Markus; Fischer, Alfons; Jacobs, Joshua; Marks, Laurence

    2012-01-01

    After the rapid growth in the use of CoCrMo metal-on-metal hip replacements since the second generation was introduced circa 1990, metal-on-metal hip replacements have experienced a sharp decline in the last two years due to biocompatibility issues related to wear and corrosion products. Despite some excellent clinical results, the release of wear and corrosion debris and the adverse response of local tissues have been of great concern. There are many unknowns regarding how CoCrMo metal bearings interact with the human body. This perspective article is intended to outline some recent progresses in understanding wear and corrosion of metal-on-metal hip replacement both in-vivo and in-vitro. The materials, mechanical deformation, corrosion, wear-assisted corrosion, and wear products will be discussed. Possible adverse health effects caused by wear products will be briefly addressed, as well as some of the many open questions such as the detailed chemistry of corrosion, tribochemical reactions and the formation of graphitic layers. Nowadays we design almost routinely for high performance materials and lubricants for automobiles; humans are at least as important. It is worth remembering that a hip implant is often the difference between walking and leading a relatively normal life, and a wheelchair. PMID:23196425

  9. CoCrMo metal-on-metal hip replacements.

    PubMed

    Liao, Yifeng; Hoffman, Emily; Wimmer, Markus; Fischer, Alfons; Jacobs, Joshua; Marks, Laurence

    2013-01-21

    After the rapid growth in the use of CoCrMo metal-on-metal hip replacements since the second generation was introduced circa 1990, metal-on-metal hip replacements have experienced a sharp decline in the last two years due to biocompatibility issues related to wear and corrosion products. Despite some excellent clinical results, the release of wear and corrosion debris and the adverse response of local tissues have been of great concern. There are many unknowns regarding how CoCrMo metal bearings interact with the human body. This perspective article is intended to outline some recent progresses in understanding wear and corrosion of metal-on-metal hip replacement both in vivo and in vitro. The materials, mechanical deformation, corrosion, wear-assisted corrosion, and wear products will be discussed. Possible adverse health effects caused by wear products will be briefly addressed, as well as some of the many open questions such as the detailed chemistry of corrosion, tribochemical reactions and the formation of graphitic layers. Nowadays we design almost routinely for high performance materials and lubricants for automobiles; humans are at least as important. It is worth remembering that a hip implant is often the difference between walking and leading a relatively normal life, and a wheelchair. PMID:23196425

  10. Total body calcium analysis. [neutron irradiation

    NASA Technical Reports Server (NTRS)

    Lewellen, T. K.; Nelp, W. B.

    1974-01-01

    A technique to quantitate total body calcium in humans is developed. Total body neutron irradiation is utilized to produce argon 37. The radio argon, which diffuses into the blood stream and is excreted through the lungs, is recovered from the exhaled breath and counted inside a proportional detector. Emphasis is placed on: (1) measurement of the rate of excretion of radio argon following total body neutron irradiation; (2) the development of the radio argon collection, purification, and counting systems; and (3) development of a patient irradiation facility using a 14 MeV neutron generator. Results and applications are discussed in detail.

  11. The stability of DLC film on nitrided CoCrMo alloy in phosphate buffer solution

    NASA Astrophysics Data System (ADS)

    Zhang, T. F.; Liu, B.; Wu, B. J.; Liu, J.; Sun, H.; Leng, Y. X.; Huang, N.

    2014-07-01

    CoCrMo alloy is often used as the material for metal artificial joint, but metal debris and metal ions are the main concern on tissue inflammation or tissue proliferation for metal prosthesis. In this paper, nitrogen ion implantation and diamond like carbon (DLC) film composite treatment was used to reduce the wear and ion release of biomedical CoCrMo substrate. The mechanical properties and stability of N-implanted/DLC composite layer in phosphate buffer solution (PBS) was evaluated to explore the full potential of N-implanted/DLC composite layer as an artificial joint surface modification material. The results showed that the DLC film on N implanted CoCrMo (N-implanted/DLC composite layer) had the higher surface hardness and wear resistance than the DLC film on virgin CoCrMo alloy, which was resulted from the strengthen effect of the N implanted layer on CoCrMo alloy. After 30 days immersion in PBS, the structure of DLC film on virgin CoCrMo or on N implanted CoCrMo had no visible change. But the adhesion and corrosion resistance of DLC on N implanted CoCrMo (N-implanted/DLC composite layer) was weakened due to the dissolution of the N implanted layer after 30 days immersion in PBS. The adhesion reduction of N-implanted/DLC composite layer was adverse for in vivo application in long term. So researcher should be cautious to use N implanted layer as an inter-layer for increasing CoCrMo alloy load carrying capacity in vivo environment.

  12. Electrochemical Testing of Ni-Cr-Mo-Gd Alloys

    SciTech Connect

    T. E. Lister; R. E. Mizia; H. Tian

    2005-10-01

    The waste package site recommendation design specified a boron-containing stainless steel, Neutronit 976/978, for fabrication of the internal baskets that will be used as a corrosion-resistant neutron-absorbing material. Recent corrosion test results gave higher-than-expected corrosion rates for this material. The material callout for these components has been changed to a Ni-Cr-Mo-Gd alloy (ASTM-B 932-04, UNS N06464) that is being developed at the Idaho National Laboratory. This report discusses the results of initial corrosion testing of this material in simulated in-package environments that could contact the fuel baskets after breach of the waste package outer barrier. The corrosion test matrix was executed using the potentiodynamic and potentiostatic electrochemical test techniques. The alloy performance shows low rates of general corrosion after initial removal of a gadolinium-rich second phase that intersects the surface. The high halide-containing test solutions exhibited greater tendencies toward initiation of crevice corrosion.

  13. Environmentally Assisted Cracking of Commercial Ni-Cr-Mo Alloys - A Review

    SciTech Connect

    Rebak, R B

    2004-11-09

    Nickel-Chromium-Molybdenum alloys (Ni-Cr-Mo) are highly resistant to general corrosion, localized corrosion and environmentally assisted cracking (EAC). Cr acts as a beneficial element under oxidizing acidic conditions and Mo under reducing conditions. All three elements (Ni, Cr and Mo) act synergistically to provide resistance to EAC in environments such as hot concentrated chloride solutions. Ni-Cr-Mo alloys may suffer EAC in environments such as hot caustic solutions, hot wet hydrofluoric acid (HF) solutions and in super critical water oxidation (SCWO) applications. Not all the Ni-Cr-Mo alloys have the same susceptibility to cracking in the mentioned environments. Most of the available data regarding EAC is for the oldest Ni-Cr-Mo alloys such as N10276 and N06625.

  14. Nonequilibrium grain-boundary cosegregation of nitrogen and chromium in NiCrMoV steel

    NASA Astrophysics Data System (ADS)

    Zheng, Lei; Xu, Tingdong

    2005-12-01

    It is concluded in this article that nonequilibrium grain-boundary cosegregation (NCGS) of nitrogen and chromium occurs in NiCrMoV steel. That conclusion is reached from experimental observations of the parallel segregation isotherms and the maximum coverage of Cr and N at grain boundaries during the isotherms. This means that the nonequilibrium segregation of Cr induces that of N, in NiCrMoV steel.

  15. DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL

    DOEpatents

    Buyers, A.G.; Rosen, F.D.; Motta, E.E.

    1959-12-22

    A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

  16. Influence of carbides and microstructure of CoCrMo alloys on their metallic dissolution resistance.

    PubMed

    Valero-Vidal, C; Casabán-Julián, L; Herraiz-Cardona, I; Igual-Muñoz, A

    2013-12-01

    CoCrMo alloys are passive and biocompatible materials widely used as joint replacements due to their good mechanical properties and corrosion resistance. Electrochemical behaviour of thermal treated CoCrMo alloys with different carbon content in their bulk alloy composition has been analysed. Both the amount of carbides in the CoCrMo alloys and the chemical composition of the simulated body fluid affect the electrochemical properties of these biomedical alloys, thus passive dissolution rate was influenced by the mentioned parameters. Lower percentage of carbon in the chemical composition of the bulk alloy and thermal treatments favour the homogenization of the surface (less amount of carbides), thus increasing the availability of Cr to form the oxide film and improving the corrosion resistance of the alloy. PMID:24094174

  17. Effect of hot deformation on phase transformation kinetics of 86CrMoV7 steel

    SciTech Connect

    Xiao Furen . E-mail: frxiao@ysu.edu.cn; Liao Bo . E-mail: cyddjyjs@263.net; Qiao Guiying; Guan Shuzhe

    2006-12-15

    The time-temperature-transformation (TTT) and continuous cooling transformation (CCT) diagrams of 86CrMoV7 steel with and without hot deformation were constructed by means of dilatometry, metallography and transmission electron microscopy (TEM). The results show that the pearlite and bainite transformations of 86CrMoV7 steel can be promoted and the microstructure can be refined by hot deformation. The undissolved carbides associated with hot deformation increase the inhomogeneity of carbon distribution in deformed austenite. The inhomogeneities of the austenite increase the number of nucleation sites for pearlite and bainite, and promote pearlite and bainite formation, which result in refinement of both the pearlite and bainite microstructures. In contrast, the undissolved carbides do not play a direct role on the pearlite and bainite transformation of 86CrMoV7 steel in the absence of hot deformation.

  18. Neoplasia in fast neutron-irradiated beagles

    SciTech Connect

    Bradley, E.W.; Zook; B.C.; Casarett, G.W.

    1981-09-01

    One hundred fifty-one beagle dogs were irradiated with either photons or fast neutrons (15 MeV) to one of three dose-limiting normal tissues - spinal cord, lung, or brain. The radiation was given in four fractions per week for 5 weeks (spinal cord), 6 weeks (lung), 7 weeks (brain) to total doses encompassing those given clinically for cancer management. To date, no nonirradiated dogs or photon-irradiated dogs have developed neoplasms within the irradiated field. Of the neutron-irradiated dogs at risk, the incidence of neoplasia was 15%. The latent period for radiation-induced cancers has varied from 1 to 4 1/2 years at this time in the study.

  19. Neoplasia in fast neutron-irradiated beagles

    SciTech Connect

    Bradley, E.W.; Zook, B.C.; Casarett, G.W.; Deye, J.A.; Adoff, L.M.; Rogers, C.C.

    1981-09-01

    One hundred fifty-one beagle dogs were irradiated with either photons or fast neutrons (15 MeV) to one of three dose-limiting normal tissues--spinal cord, lung, or brain. The radiation was given in four fractions per week for 5 weeks (spinal cord), 6 weeks (lung), or 7 weeks (brain) to total doses encompassing those given clinically for cancer management. To date, no nonirradiated dogs or photon-irradiated dogs have developed any neoplasms. Seven dogs receiving fast neutrons have developed 9 neoplasms within the irradiated field. Of the neutron-irradiated dogs at risk, the incidence of neoplasia was 15%. The latent period for radiation-induced cancers has varied from 1 to 4 1/2 years at this time in the study.

  20. Neutron irradiation induced amorphization of silicon carbide

    SciTech Connect

    Snead, L.L.; Hay, J.C.

    1998-09-01

    This paper provides the first known observation of silicon carbide fully amorphized under neutron irradiation. Both high purity single crystal hcp and high purity, highly faulted (cubic) chemically vapor deposited (CVD) SiC were irradiated at approximately 60 C to a total fast neutron fluence of 2.6 {times} 10{sup 25} n/m{sup 2}. Amorphization was seen in both materials, as evidenced by TEM, electron diffraction, and x-ray diffraction techniques. Physical properties for the amorphized single crystal material are reported including large changes in density ({minus}10.8%), elastic modulus as measured using a nanoindentation technique ({minus}45%), hardness as measured by nanoindentation ({minus}45%), and standard Vickers hardness ({minus}24%). Similar property changes are observed for the critical temperature for amorphization at this neutron dose and flux, above which amorphization is not possible, is estimated to be greater than 130 C.

  1. TEM study of neutron-irradiated iron

    SciTech Connect

    Horton, L.L.; Bentley, J.; Farrell, K.

    1981-01-01

    Results of a transmission electron microscopy study of the defect structure in iron neutron-irradiated to low fluences (less than or equal to 1 dpa) at temperatures of 455 to 1013/sup 0/K are presented. The dislocation microstructures coarsen with increasing irradiation temperature from decorated dislocations, through clusters of dislocation loops, to near-edge, interstitial dislocation loops with b = a<100>, and network segments. Significant cavity formation occurred only at 548 to 723/sup 0/K, with homogeneous distributions found only at 623 and 673/sup 0/K. The maximum swelling of 0.07% occurred at 673/sup 0/K. Large cavities had a truncated octahedral shape with (111) facets and (100) truncations. Damage halos were observed around boron-containing precipitates. The effects of interstitial impurities on microstructural development and the differences in the observed microstructures compared to those in refractory bcc metals are discussed. 8 figures, 6 tables.

  2. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  3. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  4. The susceptibility of Cr-Mo-V steel welds to radiation embrittlement

    SciTech Connect

    Ol'shanskii, N.A.; Amaev, A.D.; Gorskii, K.V.; Murav'eva, T.P.

    1983-01-01

    The effect of radiation on the susceptibility to embrittlement fracture of Cr-Mo-V steel welds, produced by electron beam, electroslag, and automatic hidden-arc methods is described. The relationship between microstructural features and susceptibility to radiation embrittlement is presented.

  5. Persistent photoconductivity in neutron irradiated GaN

    NASA Astrophysics Data System (ADS)

    Minglan, Zhang; Ruixia, Yang; Naixin, Liu; Xiaoliang, Wang

    2013-09-01

    Unintentionally doped GaN films grown by MOCVD were irradiated with neutrons at room temperature. In order to investigate the influence of neutron irradiation on the optical properties of GaN films, persistent photoconductivity (PPC) and low temperature photoluminescence (PL) measurements were carried out. Pronounced PPC was observed in the samples before and after neutron irradiation without the appearance of a yellow luminescence (YL) band in the PL spectrum, suggesting that the origin of PPC and YL are not related. Moreover, PPC phenomenon was enhanced by neutron irradiation and quenched by the followed annealing process at 900 °C. The possible origin of PPC is discussed.

  6. A separation of protactinium from neutron-irradiated thorium.

    PubMed

    Lyle, S J; Shendrikar, A D

    1966-01-01

    A convenient-method, based on liquid-liquid extraction with N-benzoyl-N-phenylhydroxylamine in chloroform, is given for the separation of protactinium-233 from neutron-irradiated thorium. PMID:18959855

  7. Neutron irradiation of human melanoma cells.

    PubMed

    Brown, K; Mountford, M H; Allen, B J; Mishima, Y; Ichihashi, M; Parsons, P

    1989-01-01

    The biological characteristics and in vitro radiosensitivity of melanoma cells to thermal neutrons were investigated as a guide to the effectiveness of boron neutron capture therapy. Plateau phase cultures of three human malignant melanoma-established cell lines were examined for cell density at confluence, doubling time, cell cycle parameters, chromosome constitution, and melanin content. Cell survival dose-response curves, for cells preincubated in the presence or absence of p-boronophenylalanine. HCl (10B1-BPA), were measured over the dose range 0.6-8.0 Gy (N + gamma). The neutron fluence rate was 2.6 x 10(9) n/cm2/s and the total dose rate 3.7 Gy/h (31% gamma). Considerable differences were observed in the morphology and cellular properties of the cell lines. Two cell lines (96E and 96L) were amelanotic, and one was melanotic (418). An enhanced killing for neutron irradiation was found only for the melanotic cells after 20 h preincubation with 10 micrograms/ml 10B1-BPA. In view of the doubling times of the cell lines of about 23 h (96E and 96L) or of 36 h (418), it seems likely that an increased boron uptake, and hence increased radiosensitivity, might result if the preincubation period with 10B1-BPA is extended to several hours longer than the respective cell cycle times. PMID:2798324

  8. PROCESSING OF NEUTRON-IRRADIATED URANIUM

    DOEpatents

    Hopkins, H.H. Jr.

    1960-09-01

    An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

  9. Phase transformations in neutron-irradiated Zircaloys

    SciTech Connect

    Chung, H.M.

    1986-04-01

    Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after approx.3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr/sub 3/O and cubic-ZrO/sub 2/ particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/)/sub 2/ and Zr/sub 2/(Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of approx.4 x 10/sup 21/ ncm/sup -2/ in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs.

  10. Weldability of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Asano, Kyoichi; Nishimura, Seiji; Saito, Yoshiaki; Sakamoto, Hiroshi; Yamada, Yuji; Kato, Takahiko; Hashimoto, Tsuneyuki

    1999-01-01

    Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10 22 to 1.4 × 10 26 n/m 2 ( E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6-16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.

  11. Fretting corrosion of CoCrMo and Ti6Al4V interfaces.

    PubMed

    Swaminathan, Viswanathan; Gilbert, Jeremy L

    2012-08-01

    Mechanically assisted corrosion (fretting corrosion, tribocorrosion etc.,) of metallic biomaterials is a primary concern for numerous implant applications, particularly in the performance of highly-loaded medical devices. While the basic underlying concepts of fretting corrosion or tribocorrosion and fretting crevice corrosion are well known, there remains a need to develop an integrated systematic method for the analysis of fretting corrosion involving metal-on-metal contacts. Such a method can provide detailed and quantitative information on the processes present and explore variations in surfaces, alloys, voltages, loadings, motion and solution conditions. This study reports on development of a fretting corrosion test system and presents elements of an in-depth theoretical fretting corrosion model that incorporates both the mechanical and the electrochemical aspects of fretting corrosion. To demonstrate the capabilities of the new system and validate the proposed model, experiments were performed to understand the effect of applied normal load on fretting corrosion performance of Ti6Al4V/Ti6Al4V, CoCrMo/Ti6Al4V, and CoCrMo/CoCrMo material couples under potentiostatic conditions with a fixed starting surface roughness. The results of this study show that fretting corrosion is affected by material couples, normal load and the motion conditions at the interface. In particular, fretting currents and coefficient of friction (COF) vary with load and are higher for Ti6Al4V/Ti6Al4V couple reaching 3 mA/cm(2) and 0.63 at about 73 MPa nominal contact stress, respectively. Ti6Al4V coupled with CoCrMo displayed lower currents (0.6 mA/cm(2)) and COF (0.3), and the fretting corrosion behavior was comparable to CoCrMo/CoCrMo couple (1.2 mA/cm(2) and 0.3, respectively). Information on the mechanical energy dissipated at the interface, the sticking behavior, and the load dependence of the inter-asperity distance calculated using the model elucidated the influence of

  12. Effect of cryogenic burnishing on surface integrity modifications of Co-Cr-Mo biomedical alloy.

    PubMed

    Yang, Shu; Dillon, Oscar W; Puleo, David A; Jawahir, Ibrahim S

    2013-01-01

    Severe plastic deformation (SPD) processes have been used to modify the surface integrity properties of many materials by generating ultrafine or even nanometer-sized grains in the surface and subsurface region. These fine grained materials created by SPD and dynamic recrystallization in a thin layer near the surface usually have higher hardness and frequently exhibit enhanced mechanical properties (wear resistance, corrosion resistance, fatigue life, etc.). Cryogenic burnishing, a SPD process, was used to improve several surface integrity parameters of a Co-Cr-Mo biomedical alloy. Application of liquid nitrogen during the burnishing process significantly suppressed the temperature rise within and outside the nitrogen application zone. Better surface finish, high hardness value, thick burnishing-influenced surface layer, and significant grain refinement were simultaneously achieved with the application of cryogenic cooling. Current results show that cryogenic burnishing can be an effective processing method for modifying the studied surface integrity properties of Co-Cr-Mo biomedical alloy. PMID:23090709

  13. Metallographic etching and microstructure characterization of NiCrMoV rotor steels for nuclear power

    NASA Astrophysics Data System (ADS)

    Liu, Peng; Lu, Feng-gui; Liu, Xia; Gao, Yu-lai

    2013-12-01

    The grain size of prior austenite has a distinct influence on the microstructure and final mechanical properties of steels. Thus, it is significant to clearly reveal the grain boundaries and therefore to precisely characterize the grain size of prior austenite. For NiCrMoV rotor steels quenched and tempered at high temperature, it is really difficult to display the grain boundaries of prior austenite clearly, which limits a further study on the correlation between the properties and the corresponding microstructure. In this paper, an effective etchant was put forward and further optimized. Experimental results indicated that this agent was effective to show the details of grain boundaries, which help analyze fatigue crack details along the propagation path. The optimized corrosion agent is successful to observe the microstructure characteristics and expected to help analyze the effect of microstructure for a further study on the mechanical properties of NiCrMoV rotor steels used in the field of nuclear power.

  14. Bovine Serum Albumin binding to CoCrMo nanoparticles and the influence on dissolution

    NASA Astrophysics Data System (ADS)

    Simoes, T. A.; Brown, A. P.; Milne, S. J.; Brydson, R. M. D.

    2015-10-01

    CoCrMo alloys exhibit good mechanical properties, excellent biocompatibility and are widely utilised in orthopaedic joint replacements. Metal-on-metal hip implant degradation leads to the release of metal ions and nanoparticles, which persist through the implant's life and could be a possible cause of health complications. This study correlates preferential binding between proteins and metal alloy nanoparticles to the alloy's corrosion behaviour and the release of metal ions. TEM images show the formation of a protein corona in all particles immersed in albumin containing solutions. Only molybdenum release was significant in these tests, suggesting high dissolution of this element when CoCrMo alloy nanoparticles are produced as wear debris in the presence of serum albumin. The same trend was observed during extended exposure of molybdenum reference nanoparticles to albumin.

  15. MODELING OF NI-CR-MO BASED ALLOYS: PART II - KINETICS

    SciTech Connect

    Turchi, P A; Kaufman, L; Liu, Z

    2006-07-07

    The CALPHAD approach is applied to kinetic studies of phase transformations and aging of prototypes of Ni-Cr-Mo-based alloys selected for waste disposal canisters in the Yucca Mountain Project (YMP). Based on a previous study on alloy stability for several candidate alloys, the thermodynamic driving forces together with a newly developed mobility database have been used to analyze diffusion-controlled transformations in these Ni-based alloys. Results on precipitation of the Ni{sub 2}Cr-ordered phase in Ni-Cr and Ni-Cr-Mo alloys, and of the complex P- and {delta}-phases in a surrogate of Alloy 22 are presented, and the output from the modeling are compared with experimental data on aging.

  16. Neutron irradiation effects in GaAs

    SciTech Connect

    Patel, J.U.

    1992-01-01

    Changes in electrical properties of n-GaAs as a result of irradiations with fast neutron have been studied, after epitaxial layers doped with Si at concentrations in the range 1.35 x 10[sup 15] to 1.60 x 10[sup 16] cm[sup [minus]3] were irradiated with reactor neutron fluences up to 1.31 x 10[sup 15] cm [sup [minus]2]. When the changes in carrier concentration, Hall mobility and resistivity were more than 25% of their initial values, nonlinear dependence on neutron fluence was apparent. New theory is proposed which explains the changes in electrical properties in terms of rates of trapping and release of charges. A theoretical relationship is derived for the change in carrier concentration as a function of neutron fluence and Fermi level shift was found to be consistent with the observed changes in carrier concentration. A correlation has been found between the changes in carrier concentration and mobility with neutron fluence using newly defined physically meaningful parameters in the case of two pairs of samples. The correlation has been explained in terms of the increased scattering of charge carriers from the defects created by neutrons that trap the free carriers. Mobility changes were measured at temperatures from 15 K to 305 K in n-GaAs van-der Pauw samples irradiated by fast reactor neutrons. The inverse mobility values obtain versus temperature, from the variable temperature Hall measurements, in the case of irradiated and in-irradiated samples were fitted using the relation [mu][sup [minus]1] = T[sup [minus]3/2] + B T[sup 3/2]. The inverse mobility increased as a result of neutron irradiations over the whole range of temperature, the increase being attributed to the increased scattering from neutron induced charged defects.

  17. Mechanism and kinetics of interaction of Fe, Cr, Mo, and Mn atoms with molecular oxygen

    SciTech Connect

    Akhmadov, U.S.; Zaslonko, I.S.; Smirnov, V.N.

    1988-09-01

    By means of resonance atomic absorption in shock waves, rate constants have been measured for the interaction of atoms of a number of transition metals (Fe, Cr, Mo, and Mn) with molecular oxygen. A new method is proposed and used for determining the exponent ..gamma.. in the modified Lambert-Beer law D = element of(ZN)/sup ..gamma../. The bond strength in CrO and MoO molecules has been estimated.

  18. Control of surface morphology of carbide coating on Co-Cr-Mo implant alloy.

    PubMed

    Vandamme, N S; Topoleski, L D T

    2005-07-01

    Wear of materials used in artificial joints is a common failure mode of artificial joints. A low wear rate for implants is believed to be critical for extending implant service time. We developed a carbide-coated Co-Cr-Mo implant alloy created in plasma of methane and hydrogen mixed gas by a microwave plasma-assisted surface reaction. The carbide-coated Co-Cr-Mo has a unique "brain coral-like" surface morphology and is much harder than uncoated Co-Cr-Mo. The effect of plasma processing time and temperature on the surface morphology of the top carbide layer was studied toward optimizing the surface coating. The ratios of average roughness, Ra, core roughness, Rk, and summation of core roughness, reduced peak height (Rpk) and reduced valley depth (Rvk), Rk+Rpk+Rvk, for the 6-h/985 degrees C coating to those for the 0.5-h/985 degrees C coating were 1.9, 1.7, and 1.9, respectively. The ratios of Ra, Rk, and Rk+Rpk+Rvk for the 4-h/1000 degrees C coating to those for the 4-h/939 degrees C coating were 2.3, 2.3, and 2.0, respectively. With the proper combination of plasma processing time and temperature, it may be possible to change the thickness of the peak-valley top cluster by fourfold from approximately 0.6 microm to approximately 2.5 microm. Finally, the growth mechanism of the carbide layers on Co-Cr-Mo was discussed in the context of atomic composition analysis. PMID:15965597

  19. [Use of powder metallurgy for development of implants of Co-Cr-Mo alloy powder].

    PubMed

    Dabrowski, J R

    2001-04-01

    This paper discusses the application of powder metallurgy for the development of porous implantation materials. Powders obtained from Co-Cr-Mo alloy with different carbon content by water spraying and grinding, have been investigated. Cold pressing and rotary re-pressing methods were used for compressing the powder. It was found that the sintered materials obtained from water spraying have the most advantageous properties. PMID:11388037

  20. Magnetism in Sr2CrMoO6 : A combined ab initio and model study

    NASA Astrophysics Data System (ADS)

    Sanyal, Prabuddha; Halder, Anita; Si, Liang; Wallerberger, Markus; Held, Karsten; Saha-Dasgupta, Tanusri

    2016-07-01

    Using a combination of first-principles density functional theory (DFT) calculations and exact diagonalization studies of a first-principles derived model, we carry out a microscopic analysis of the magnetic properties of the half-metallic double perovskite compound Sr2CrMoO6 , a sister compound of the much discussed material Sr2FeMoO6 . The electronic structure of Sr2CrMoO6 , though appearing similar to Sr2FeMoO6 at first glance, shows nontrivial differences with that of Sr2FeMoO6 on closer examination. In this context, our study highlights the importance of charge transfer energy between the two transition metal sites. The change in charge transfer energy due to a shift of Cr d states in Sr2CrMoO6 compared to Fe d in Sr2FeMoO6 suppresses the hybridization between Cr t2 g and Mo t2 g. This strongly weakens the hybridization-driven mechanism of magnetism discussed for Sr2FeMoO6 . Our study reveals that, nonetheless, the magnetic transition temperature of Sr2CrMoO6 remains high since an additional superexchange contribution to magnetism arises with a finite intrinsic moment developed at the Mo site. We further discuss the situation in comparison to another related double perovskite compound, Sr2CrWO6 . We also examine the effect of correlation beyond DFT, using dynamical mean field theory.

  1. Matched-pair total knee arthroplasty retrieval analysis: oxidized zirconium vs. CoCrMo.

    PubMed

    Heyse, Thomas J; Chen, Dan X; Kelly, Natalie; Boettner, Friedrich; Wright, Timothy M; Haas, Steven B

    2011-12-01

    Oxidized zirconium (OxZr) was introduced to serve as a ceramic surface for femoral components in TKA. The aim of this study was to compare retrieved OxZr components and corresponding PE inserts in matched comparison with conventional cobalt/chrome/molybdenum alloy (CoCrMo). Eleven retrieved posterior stabilized TKA with an OxZr femoral component were included. This included 6 implants from an earlier preliminary study. From a cohort of 56 retrieved TKA with conventional CoCrMo femoral components, pairs were matched according to duration of implantation, patient age, reason for revision, and BMI. Polyethylene inlays and femoral components were optically scored for in vivo damage. The average damage score of the tibial PE inserts was significantly lower with OxZr components (p=0.01). Mainly burnishing and scratches were found. The average wear score in the visual analysis of the femoral components was significantly lower for the OxZr as well (p=0.005). Femoral components made of OxZr were less sensitive to in vivo damage and corresponding PE inlays also showed less damage than CoCrMo components. PMID:20869251

  2. Precipitation behavior of BN type inclusions in 42CrMo steel

    NASA Astrophysics Data System (ADS)

    Wang, Yu-nan; Bao, Yan-ping; Wang, Min; Zhang, Le-chen

    2013-01-01

    Automobile crankshaft steel 42CrMo, which requires excellent machinability and mechanical properties, cannot be manufactured by traditional methods. To achieve these qualities, the formation behavior of boron nitride (BN) inclusions in 42CrMo steel was studied in this article. First, the precipitation temperature and the amount of BN type inclusions with different contents of boron and nitrogen in molten steel were calculated thermodynamically by FactSage software. Then the morphology and the size of BN type inclusions as well as the influence of cooling methods on them were investigated by scanning electron microscopy. Furthermore, the effects of cooling rate and the contents of B and N in molten steel on the morphology, size, and distribution of BN type inclusions were studied quantitatively and detailedly by directional solidification experiments. It is found that different BN inclusions in molten steel can form by controlling the cooling rate and the contents of B and N, which is important for obtaining the excellent machinability of 42CrMo steel.

  3. Assessment of precipitation behavior in dental castings of a Co-Cr-Mo alloy.

    PubMed

    Yamanaka, Kenta; Mori, Manami; Chiba, Akihiko

    2015-10-01

    This study investigated solute portioning and precipitation in dental castings of a Co-Cr-Mo alloy and discussed their effects on alloy performance, in particular, the mechanical properties. Samples of a commercial Co-29Cr-6Mo (mass%) alloy were prepared using a dental-casting machine. The precipitates formed owing to the partitioning behaviors of the alloying elements were investigated using scanning electron microscopy, electron backscatter diffraction analysis, electron probe microanalysis, and transmission electron microscopy. The prepared samples exhibited a very coarse face-centered-cubic γ-phase dendritic structure with an average grain size of a few millimeters. A large number of precipitates, which decomposed further into complex interdendritic constituents (σ- and M23C6 carbide phases) were observed in the interdendritic regions rich in Cr, Mo, Si, and C. A reaction between the σ-phase and carbon is probably responsible for the carbide M23C6; however, this reaction did not occur to completion in the current case in spite of slow cooling (i.e., long exposure to elevated temperatures) in dental casting. While these precipitates result in high strength (hardness) and/or brittleness, the properties can be improved further by optimizing the alloy composition and the manufacturing process. The results of this study shed light on the significance of precipitation control in dental castings of Co-Cr-Mo alloys and should aid in the design of novel biomedical Co-Cr-based dental alloys that exhibit better performances. PMID:26164217

  4. Acid etching does not improve CoCrMo implant osseointegration in a canine implant model.

    PubMed

    Jakobsen, Stig S; Baas, Jorgen; Jakobsen, Thomas; Soballe, Kjeld

    2010-01-01

    Induction of bone ingrowth by topographical changes to implant surfaces is an attractive concept. Topographical modifications achieved by acid etching are potentially applicable to complex 3D surfaces. Using clinically relevant implant models, we explored the effect of wet etching porous bead-coated CoCrMo. The study was designed as two paired animal experiments with 10 dogs. Each dog received four implants; one in each medial femoral condyle (loaded 0.75-mm-gap model) and one in each proximal tibia (press-fit). The implants were observed for 6 weeks and were evaluated by biomechanical pushout tests and histomorphometry. We found that wet etching porous bead-coated CoCrMo implants failed to improve implant performance. Moreover, a tendency towards increased fibrous tissue formation, decreased new bone formation, and decreased mechanical fixation was observed. Surface topography on implants is able to stimulate bone-forming cells, but the clinical performance of an implant surface perhaps relies more on 3D geometrical structure and biocompatibility. Caution should be exercised regarding the results of wet etching of porous bead-coated CoCrMo and there is a need for more preclinical trials. PMID:20544657

  5. Proton and neutron irradiation effect of Ti: Sapphires

    SciTech Connect

    Wang, G.; Zhang, J.; Yang, J.

    1999-07-01

    Various effects of proton and neutron irradiated Ti: sapphires were studied. Proton irradiation induced F, F{sup +} and V center in Ti: sapphires and 3310 cm{sup -1} infrared absorption, and made ultraviolet absorption edge shift to short wave. Neutron irradiation produced a number of F, F{sup +} and F{sub 2} centers and larger defects in Ti: sapphires, and changed Ti{sup 4+}into Ti{sup 3+} ions. Such valence state variation enhanced characteristic luminescence of Ti: sapphires, and no singular variances of intrinsic fluorescence spectra of Ti: sapphires took place with neutron flux of 1 x 10{sup 17}n/cm{sup 2}, but the fluorescence vanished with neutron flux of 1 x 10{sup 18}n/cm{sup 2} which means the threshold for the concentration of improving Ti{sup 3+} ions by neutron irradiation.

  6. New facility for post irradiation examination of neutron irradiated beryllium

    SciTech Connect

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-09-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800{degrees}C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and {sup 60}Co;7.4 MBq/day.

  7. Laser annealing of neutron irradiated boron-10 isotope doped diamond

    SciTech Connect

    Jagannadham, K.; Butler, J. E.

    2011-01-01

    10B isotope doped p-type diamond epilayer grown by chemical vapor deposition on (110) oriented type IIa diamond single crystal substrate was subjected to neutron transmutation at a fluence of 2.4 9 1020 thermal and 2.4 9 1020 fast neutrons. After neutron irradiation, the epilayer and the diamond substrate were laser annealed using Nd YAG laser irradiation with wave length, 266 nm and energy, 150 mJ per pulse. The neutron irradiated diamond epilayer and the substrate were characterized before and after laser annealing using different techniques. The characterization techniques include optical microscopy, secondary ion mass spectrometry, X-ray diffraction, Raman, photoluminescence and Fourier Transform Infrared spectroscopy, and electrical sheet conductance measurement. The results indicate that the structure of the irradiation induced amorphous epilayer changes to disordered graphite upon laser annealing. The irradiated substrate retains the (110) crystalline structure with neutron irradiation induced defects.

  8. Potential and frequency effects on fretting corrosion of Ti6Al4V and CoCrMo surfaces.

    PubMed

    Swaminathan, Viswanathan; Gilbert, Jeremy L

    2013-09-01

    Fretting corrosion has been reported at the metal-metal interfaces of a wide range of medical devices, including total joint replacements, spinal devices, and overlapping cardiovascular stents. Currently, the fretting corrosion phenomenon associated with metal-on-metal contacts is not fully understood. This study investigated the effect of potential and fretting frequency on the fretting corrosion performance of Ti6Al4V/Ti6Al4V, Ti6Al4V/CoCrMo, and CoCrMo/CoCrMo alloy combinations at fixed normal load and displacement conditions using a custom built fretting corrosion test system. The results showed that the fretting current densities increased with increases in potential and were highest for Ti6Al4V/Ti6Al4V couple (1.5 mA/cm(2) at 0 V vs. Ag/AgCl). The coefficient of friction varied with potential and was about two times higher for Ti6Al4V/Ti6Al4V (0.71 V at 0 V vs. Ag/AgCl). In most of the potential range tested, the fretting corrosion behavior of CoCrMo/Ti6Al4V and CoCrMo/CoCrMo was similar and dominated by the CoCrMo surface. Increase in applied fretting frequency linearly increased the fretting current densities in the regions where the passive film is stable. Also, the model-based fretting current densities were in excellent agreement with the experimental results. Overall, Ti6Al4V/Ti6Al4V couple was more susceptible to fretting corrosion compared with other couples. However, the effects of these processes on the biological system were not assessed. PMID:23404905

  9. Mesenchymal stem cell response to topographically modified CoCrMo

    PubMed Central

    Logan, Niall; Bozec, Laurent; Traynor, Alison

    2015-01-01

    Abstract Surface roughness on implant materials has been shown to be highly influential on the behavior of osteogenic cells. Four surface topographies were engineered on cobalt chromium molybdenum (CoCrMo) in order to examine this influence on human mesenchymal stem cells (MSC). These treatments were smooth polished (SMO), acid etched (AE) using HCl 7.4% and H2SO4 76% followed by HNO3 30%, sand blasted, and acid etched using either 50 μm Al2O3 (SLA50) or 250 μm Al2O3 grit (SLA250). Characterization of the surfaces included energy dispersive X‐ray analysis (EDX), contact angle, and surface roughness analysis. Human MSCs were cultured onto the four CoCrMo substrates and markers of cell attachment, retention, proliferation, cytotoxicity, and osteogenic differentiation were studied. Residual aluminum was observed on both SLA surfaces although this appeared to be more widely spread on SLA50, whilst SLA250 was shown to have the roughest topography with an R a value greater than 1 μm. All substrates were shown to be largely non‐cytotoxic although both SLA surfaces were shown to reduce cell attachment, whilst SLA50 also delayed cell proliferation. In contrast, SLA250 stimulated a good rate of proliferation resulting in the largest cell population by day 21. In addition, SLA250 stimulated enhanced cell retention, calcium deposition, and hydroxyapatite formation compared to SMO (p < 0.05). The enhanced response stimulated by SLA250 surface modification may prove advantageous for increasing the bioactivity of implants formed of CoCrMo. © 2015 Wiley Periodicals, Inc. J Biomed Mater Res Part A: 103A: 3747–3756, 2015. PMID:26015290

  10. Mesenchymal stem cell response to topographically modified CoCrMo.

    PubMed

    Logan, Niall; Bozec, Laurent; Traynor, Alison; Brett, Peter

    2015-12-01

    Surface roughness on implant materials has been shown to be highly influential on the behavior of osteogenic cells. Four surface topographies were engineered on cobalt chromium molybdenum (CoCrMo) in order to examine this influence on human mesenchymal stem cells (MSC). These treatments were smooth polished (SMO), acid etched (AE) using HCl 7.4% and H2SO4 76% followed by HNO3 30%, sand blasted, and acid etched using either 50 μm Al2O3 (SLA50) or 250 μm Al2 O3 grit (SLA250). Characterization of the surfaces included energy dispersive X-ray analysis (EDX), contact angle, and surface roughness analysis. Human MSCs were cultured onto the four CoCrMo substrates and markers of cell attachment, retention, proliferation, cytotoxicity, and osteogenic differentiation were studied. Residual aluminum was observed on both SLA surfaces although this appeared to be more widely spread on SLA50, whilst SLA250 was shown to have the roughest topography with an Ra value greater than 1 μm. All substrates were shown to be largely non-cytotoxic although both SLA surfaces were shown to reduce cell attachment, whilst SLA50 also delayed cell proliferation. In contrast, SLA250 stimulated a good rate of proliferation resulting in the largest cell population by day 21. In addition, SLA250 stimulated enhanced cell retention, calcium deposition, and hydroxyapatite formation compared to SMO (p < 0.05). The enhanced response stimulated by SLA250 surface modification may prove advantageous for increasing the bioactivity of implants formed of CoCrMo. PMID:26015290

  11. Corrosion and degradation of a polyurethane/Co-Ni-Cr-Mo pacemaker lead

    SciTech Connect

    Sung, P.; Fraker, A.C.

    1987-12-01

    An investigation to study changes in the metal surfaces and the polyurethane insulation of heart pacemaker leads under controlled in vitro conditions was conducted. A polyurethane (Pellethane 2363-80A)/Co-Ni-Cr-Mo (MP35N) wire lead was exposed in Hanks' physiological saline solution for 14 months and then analyzed using scanning electron microscopy, x-ray energy dispersive analysis, and small angle x-ray scattering. Results showed that some leakage of solution into the lead had occurred and changes were present on both the metal and the polyurethane surfaces.

  12. A study of the 42CrMo4 steel surface by quantitative XPS electron spectroscopy

    NASA Astrophysics Data System (ADS)

    Flori, M.; Gruzza, B.; Bideux, L.; Monier, G.; Robert-Goumet, C.

    2008-05-01

    Quantitative X-ray photoelectron spectroscopy was used to characterize the native oxide film formed on 42CrMo4 steel surface by air exposure in normal conditions. In order to determine the thickness and composition of the oxide layer we have used a stacking layer model together with experimental XPS sputtering depth profiling. At a nanoscale study, to obtain quantitative results one must take into account fundamental parameters like the attenuation depth of photoelectrons. We have found that both lepidocrocit (γ-FeOOH) and magnetite (Fe 3O 4) were present and the total thickness of the oxide layer was 16 monolayers.

  13. Bright nitriding of Cr-Mo-steels in plasma and gas

    SciTech Connect

    Larisch, B.; Spies, H.J.; Hoeck, K.

    1995-12-31

    Although the reduction of the white layer in special gas atmospheres directly after nitriding and bright nitriding were reported a long time ago, the white layer is mostly removed by mechanical or chemical means in industrial practice. The main reason for this is poor process control. However, new requirements such as the duplex treatment (nitriding + hardcoating), demand a more detailed examination of bright nitriding. Today, new possibilities exist for process control in gas nitriding by solid electrolyte sensors. Steel grades 17CrMoV10 and 31CrMoV9 were bright nitrided in gas and plasma. In contrast to the above experiments, in the two-step technology no white layer forms in the first step (20min) at a higher nitriding potential. By this, the formation of a soft surface layer (of iron) can be avoided. Limits of this technology--for instance in the depth of the formed nitrided case--are discussed. Reasons for the often discussed faster nitriding in plasma are explained on the basis of the experimental results. The influence of ion bombardment in plasma nitriding on the activation of the surface and the nitriding results is discussed in comparison to gas nitriding. In this context the advantages of plasma nitriding--with respect to higher chromium alloyed steels (>5%Cr), which tend to passivation--are shown.

  14. Reduction of Cr, Mo, Se and U by Desulfovibrio desulfuricans immobilized in polyacrylamide gels.

    PubMed

    Tucker, M D; Barton, L L; Thomson, B M

    1998-01-01

    Intact cells of Desulfovibrio desulfuricans, immobilized in polyacrylamide gel, removed Cr, Mo, Se and U from solution by enzymatic-mediated reduction reactions. Lactate or H2 served as the electron donor and the oxidized Cr(VI), Mo(VI), Se(VI) and U(VI) served as electron acceptors. Reduction of the oxidized metal species resulted in the precipitation of solid phases of the metals. Metal removal efficiencies of 86-96% were achieved for initial concentrations of 1 mM Mo, Se, and U and 0.5 mM Cr. Insoluble metal phases accumulated on both the surface and the interior of the polyacrylamide gel. In column tests conducted for U removal, effluent concentrations less than 20 micrograms L(-1) were achieved with initial concentrations of 5 mg L(-1) and 20 mg L(-1) U and residence times from 25-37 h. The enzymatic reduction of Cr, Mo, Se, and U by immobilized cells of D. desulfuricans may be a practical method for removing these metals from solution in a biological reactor. PMID:9565467

  15. Improving Cleanliness of 95CrMo Drill Rod Steel by Slag Refining

    NASA Astrophysics Data System (ADS)

    Wang, Linzhu; Yang, Shufeng; Li, Jingshe; Wu, Tuo; Liu, Wei; Xiong, Jiaze

    2016-02-01

    Industrial experiments were performed to improve the cleanliness of 95CrMo drill rod steel by slag refining. Higher steel cleanliness, lower corrosion, and small inclusions were obtained using the optimal slag composition (pctCaO/pctSiO2 = 3.7 to 4, pctCaO/pctAl2O3 = 6 to 8). Layered composite inclusions formed during vacuum decarburizing refining. CaS first precipitated around the spinel and subsequently formed inclusions in which solid CaS-CaO wrapped around the Al2O3-MgO-SiO2-CaO system as the modification and diffusion progressed. The thermodynamic equilibrium between slag and liquid 95CrMo steel at 1873 K (1600 °C) was also studied to understand the effect of slag composition on the oxygen content and absorption capacity for Al2O3. A mathematical model based on an investigation of slag viscosity and the interfacial tension between slag and inclusions was used to predict the size of critical inclusions for different slags. The evolution of typical inclusions is discussed in terms of the study of reactions between slag and steel.

  16. Nanostructure characterisation of flow-formed Cr-Mo-V steel using transmission Kikuchi diffraction technique.

    PubMed

    Birosca, S; Ding, R; Ooi, S; Buckingham, R; Coleman, C; Dicks, K

    2015-06-01

    Nowadays flow-forming has become a desired near net shape manufacturing method as it provides excellent mechanical properties with improved surface finish and significant manufacturing cost reduction. However, the material is subjected to excessive plastic deformation during flow-forming process, generating a very fine and complex microstructure. In addition, the intense dislocation density and residual stress that is generated in the component during processing makes the microstructure characterisation using conventional micro-analytical tools challenging. Thus, the microstructure/property relationship study in such a material is rather difficult. In the present study a flow-formed Cr-Mo-V steel nanostructure and crystallographic texture were characterised by means of Transmission Kikuchi Diffraction (TKD). Here, TKD is shown to be a powerful technique in revealing very fine martensite laths within an austenite matrix. Moreover, fine precipitates in the order of 20-70 nm on the martensite lath boundaries were clearly imaged and characterised. This greatly assisted in understanding the preferable site formation of the carbides in such a complex microstructure. The results showed that the actual TKD spatial resolution was in the range of 5-10 nm using 25 kV for flow-formed Cr-Mo-V steel. PMID:25697460

  17. Biomaterial Co-Cr-Mo Alloys Nano Coating Calcium Phosphate Orthopedic Treatment

    NASA Astrophysics Data System (ADS)

    Palaniappan, N.; Inwati, Gajendra Kumar; Singh, Man

    2014-08-01

    The modem study a thermal martensitic transformation of biomedical Co-Cr-Mo alloys and ultimately offers large elongation to failure while maintaining high strength. In the future study, structural evolution and dislocation slip as an elementary process in the martensitic transformation in Co-Cr-Mo alloys were investigated to reveal the origin of their enhanced phase stability due to nitrogen addition and coating of calcium phosphate specimens with and without nitrogen addition were prepared. The N-doped alloys had a single-phase matrix, whereas the N-free alloys had a duplex microstructure. Irrespective of the nitrogen content, dislocations frequently dissociated into Shockley partial dislocations with stacking faults. The Nano range coating of calcium phosphate function as obstacles to the glide of partial dislocations and consequently significantly affect the kinetics of the martensitic transformation. As a result, the formation of marten site plays a crucial role in plastic deformation and wear behavior, the developed nanostructures modification associated with nitrogen addition must be a promising strategy for highly durable orthopedic implants.

  18. Investigation on Mechanical Properties of 9%Cr/CrMoV Dissimilar Steels Welded Joint

    NASA Astrophysics Data System (ADS)

    Liu, Xia; Lu, Fenggui; Yang, Renjie; Wang, Peng; Xu, Xiaojin; Huo, Xin

    2015-04-01

    Advanced 9%Cr steel with good heat resistance and CrMoV with good toughness were chosen as candidate materials to fabricate combined rotor for steam turbine operating at over 620 °C. But the great difference in base metals properties presents a challenge in achieving sound defect-free joint with optimal properties in dissimilar welded rotor. In this paper, appropriate selection of filler metal, welding parameters, and post-weld heat treatment was combined to successfully weld 1100-mm-diameter 9%Cr/CrMoV dissimilar experimental rotor through ultra-narrow gap submerge arc welding. Some properties such as hardness, low-cycle fatigue (LCF), and high-cycle fatigue (HCF) combined with microstructural characterization qualify the integrity of the weld. Microstructural analysis indicated the presence of high-temperature tempered martensite as the phase responsible for the improved properties obtained in the weld. The Coffin-Manson parameters were obtained by fitting the data in LCF test, while the conditional fatigue strength was derived from the HCF test based on S-N curve. Analysis of hardness profile showed that the lowest value occurred at heat-affected zone adjacent to base metal which represents the appropriate location of fracture for the samples after LCF and HCF tests.

  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. Abrasion resistance of oxidized zirconium in comparison with CoCrMo and titanium nitride coatings for artificial knee joints.

    PubMed

    Galetz, Mathias C; Fleischmann, Ernst W; Konrad, Christian H; Schuetz, Adelheid; Glatzel, Uwe

    2010-04-01

    Most total knee replacement joints consist of a metal femoral component made from a cobalt-chromium- molybdenum (CoCrMo)-alloy and a tibial component with an ultrahigh molecular weight polyethylene (UHMWPE) bearing surface. Wear of the UHMWPE remains the primary disadvantage of these implants. The allergic potential ascribed to CoCrMo-alloys is a further concern. Other metallic alloys with and without ceramic coatings are clinically used to avoid these problems. This study compared the mechanical surface properties of an oxidized zirconium alloy with those of cast and wrought CoCrMo and TiAlV6-4. Additionally, the influence of a titanium nitride (TiN)-plasma coating on the surface properties was investigated. The composition of the oxidized zirconium layer was analyzed. Micro- and macrohardness tests as well as adhesion tests were used to reveal material differences in terms of their abrasive wear potential in artificial joints. PMID:20162723

  1. Effect of bimodal harmonic structure design on the deformation behaviour and mechanical properties of Co-Cr-Mo alloy.

    PubMed

    Vajpai, Sanjay Kumar; Sawangrat, Choncharoen; Yamaguchi, Osamu; Ciuca, Octav Paul; Ameyama, Kei

    2016-01-01

    In the present work, Co-Cr-Mo alloy compacts with a unique bimodal microstructural design, harmonic structure design, were successfully prepared via a powder metallurgy route consisting of controlled mechanical milling of pre-alloyed powders followed by spark plasma sintering. The harmonic structured Co-Cr-Mo alloy with bimodal grain size distribution exhibited relatively higher strength together with higher ductility as compared to the coarse-grained specimens. The harmonic Co-Cr-Mo alloy exhibited a very complex deformation behavior wherein it was found that the higher strength and the high retained ductility are derived from fine-grained shell and coarse-grained core regions, respectively. Finally, it was observed that the peculiar spatial/topological arrangement of stronger fine-grained and ductile coarse-grained regions in the harmonic structure promotes uniformity of strain distribution, leading to improved mechanical properties by suppressing the localized plastic deformation during straining. PMID:26478398

  2. 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.

  3. Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

    SciTech Connect

    Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu; Mohamed, Abdul Aziz; Karim, Julia Abdul

    2011-03-30

    Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO{sub 4}) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1x10{sup 13} ncm{sup -2}s{sup -1}. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emission of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.

  4. Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

    NASA Astrophysics Data System (ADS)

    Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu; Mohamed, Abdul Aziz; Karim, Julia Abdul

    2011-03-01

    Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO4) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1×1013 ncm-2s-1. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emission of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.

  5. Neutron irradiation influence on magnesium aluminium spinel inversion

    NASA Astrophysics Data System (ADS)

    Skvortsova, V.; Mironova-Ulmane, N.; Ulmanis, U.

    2002-05-01

    Grown by the Verneuil method MgO · nAl 2O 3 single crystals and natural spinel crystal have been studied using X-ray diffraction and photoluminescence spectra. The fast neutron irradiation of magnesium aluminium spinel leads to the lattice parameter decrease. The bond lengths of Mg-O and Al-O vary with the u-parameter and the lattice parameter. On the other hand, the bond lengths are related with the inversion parameter. Using changes of the lattice parameter during irradiation we have calculated the inversion parameter, which is 15-20%. In the luminescence spectra, the fast neutron radiation (fluence 10 16 cm -2) produces an increase in the intensity ratio of the N- to R-lines by 5-20%. Taking into account that intensity of the N-lines is closely associated with the inversion parameter, it is possible to state that the neutron irradiation causes the increasing of the spinel inversion.

  6. Evaluation of weld crack susceptibility for neutron irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Suzuki, T.; Kohyama, A.; Hirose, T.; Narui, M.

    In order to clarify the mechanisms of weld cracking, especially for heat affected zone cracking in heavily neutron irradiated stainless steels and to establish a measure to evaluate crack susceptibility, a mini-sized Varestraint (variable restraint) test machine for hot laboratory operation was designed and fabricated. This unique PIE facility was successfully applied in the hot laboratory of IMR Oarai Branch of Tohoku University. The maximum restraint applied was 4% at the surface of the specimen. Specimen surface morphology and specimen microstructures were inspected by video microscope, SEM and TEM. Under the 2% surface restraint condition, clear formation of heat affected zone (HAZ) crack was observed for the case of neutron irradiation to produce 0.5 appm He and of 2.4 kJ heat input by TIG.

  7. Neutron irradiation effects on high Nicalon silicon carbide fibers

    SciTech Connect

    Osborne, M.C.; Steiner, D.; Snead, L.L.

    1996-10-01

    The effects of neutron irradiation on the mechanical properties and microstructure of SiC and SiC-based fibers is a current focal point for the development of radiation damage resistant SiC/SiC composites. This report discusses the radiation effects on the Nippon Carbon Hi-Nicalon{trademark} fiber system and also discusses an erratum on earlier results published by the authors on this material. The radiation matrix currently under study is also summarized.

  8. 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.

  9. Evaluation of Neutron Irradiated Silicon Carbide and Silicon Carbide Composites

    SciTech Connect

    Newsome G, Snead L, Hinoki T, Katoh Y, Peters D

    2007-03-26

    The effects of fast neutron irradiation on SiC and SiC composites have been studied. The materials used were chemical vapor deposition (CVD) SiC and SiC/SiC composites reinforced with either Hi-Nicalon{trademark} Type-S, Hi-Nicalon{trademark} or Sylramic{trademark} fibers fabricated by chemical vapor infiltration. Statistically significant numbers of flexural samples were irradiated up to 4.6 x 10{sup 25} n/m{sup 2} (E>0.1 MeV) at 300, 500 and 800 C in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Dimensions and weights of the flexural bars were measured before and after the neutron irradiation. Mechanical properties were evaluated by four point flexural testing. Volume increase was seen for all bend bars following neutron irradiation. Magnitude of swelling depended on irradiation temperature and material, while it was nearly independent of irradiation fluence over the fluence range studied. Flexural strength of CVD SiC increased following irradiation depending on irradiation temperature. Over the temperature range studied, no significant degradation in mechanical properties was seen for composites fabricated with Hi-Nicalon{trademark} Type-S, while composites reinforced with Hi-Nicalon{trademark} or Sylramic fibers showed significant degradation. The effects of irradiation on the Weibull failure statistics are also presented suggesting a reduction in the Weibull modulus upon irradiation. The cause of this potential reduction is not known.

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

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

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

  11. Exceptionally high glass-forming ability of an FeCoCrMoCBY alloy

    SciTech Connect

    Shen Jun; Chen Qingjun; Sun Jianfei; Fan Hongbo; Wang Gang

    2005-04-11

    It has been well documented that the maximum thickness of as-cast glassy samples attainable through conventional metallurgical routes is the decisive criteria for measuring the glass-forming ability (GFA) of bulk metallic glasses (BMGs). Here we report the exceptionally high GFA of an FeCoCrMoCBY alloy which can be fabricated in the form of glassy rods with a maximum sample thickness of at least 16 mm. It is demonstrated that, by substituting Fe with a proper amount of Co in a previously reported Fe-based BMG alloy, the glass formation of the resultant new alloy can be extensively favored both thermodynamically and kinetically. The new ferrous BMG alloy also exhibits a high fracture strength of 3500 MPa and Vickers hardness of 1253 kg mm{sup -2}.

  12. Ratcheting induced cyclic softening behaviour of 42CrMo4 steel

    NASA Astrophysics Data System (ADS)

    Kreethi, R.; Mondal, A. K.; Dutta, K.

    2015-02-01

    Ratcheting is an important field of fatigue deformation which happens under stress controlled cyclic loading of materials. The aim of this investigation is to study the uniaxial ratcheting behavior of 42CrMo4 steel in annealed condition, under various applied stresses. In view of this, stress controlled fatigue tests were carried out at room temperature up to 200 cycles using a servo-hydraulic universal testing machine. The results indicate that accumulation of ratcheting strain increases monotonically with increasing maximum applied stress however; the rate of strain accumulation attains a saturation plateau after few cycles. The investigated steel shows cyclic softening behaviour under the applied stress conditions. The nature of strain accumulation and cyclic softening has been discussed in terms of dislocation distribution and plastic damage incurred in the material.

  13. Entropy and Diffuse Scattering: Comparison of NbTiVZr and CrMoNbV

    NASA Astrophysics Data System (ADS)

    Widom, Michael

    2016-07-01

    The chemical disorder intrinsic to high-entropy alloys inevitably creates diffuse scattering in their X-ray or neutron diffraction patterns. Through first principles hybrid Monte Carlo/molecular dynamics simulations of two BCC high-entropy alloy forming compounds, CrMoNbV and NbTiVZr, we identify the contributions of chemical disorder, atomic size, and thermal fluctuations to the diffuse scattering. As a side benefit, we evaluate the reduction in entropy due to pair correlations within the framework of the cluster variation method. Finally, we note that the preference of Ti and Zr for hexagonal structures at low temperature leads to a mechanical instability reducing the local BCC character of NbTiVZr, while preserving global BCC symmetry.

  14. Carbide precipitation, grain boundary segregation, and temper embrittlement in NiCrMoV rotor steels

    NASA Astrophysics Data System (ADS)

    Bandyopadhyay, N.; Briant, C. L.; Hall, E. L.

    1985-05-01

    This paper presents a study of carbide precipitation, grain boundary segregation, and temper embrittlement in NiCrMoV rotor steels. One of the steels was high purity, one was doped with phosphorus, one was doped with tin, and one was commercial purity. In addition, two NiCrV steels, one high purity and one doped with phosphorus, were examined. Carbide precipitation was studied with analytical electron microscopy. It was found that after one hour of tempering at 600 ‡C only M3C carbides were precipitated in the NiCrMoV steels. These were very rich in iron. As the tempering time increased, the chromium content of the M3C carbides increased significantly, but their size did not change. Chromium rich M7C3 precipitates began to form after 20 hours of tempering, and after 50 hours of tempering Mo-rich M2C carbides were precipitated. Also, after 100 hours of tempering, the matrix formed bands rich in M3C or M7C3 and M2C particles. Tempering occurred more rapidly in the NiCrV steels. Grain boundary segregation was studied with Auger electron spectroscopy. It was found that the amount of phosphorus and tin segregation that occurred during a step-cooling heat treatment after tempering was less if a short time tempering treatment had been used. It will be proposed that this result occurs because the low temperature tempering treatments leave more carbon in the matrix. Carbon then compctes with phosphorus and tin for sites at grain boundaries. This compctition appears to affect phosphorus segregation more than tin segregation. In addition to these two impurity elements, molybdenum and nickel segregated during low temperature aging. The presence of molybdenum in the steel did not appear to affect phosphorus segregation. Finally, it will be shown that all of the steels that contain phosphorus and/or tin exhibit some degree of temper embrittlement when they are aged at 520 ‡C or are given a step-cooling heat treatment. Of the NiCrMoV steels, the phosphorus-doped steel showed

  15. Noncontact Evaluation of Surface-Wave Nonlinearity for Creep Damage in Cr-Mo-V Steel

    NASA Astrophysics Data System (ADS)

    Ohtani, Toshihiro; Ogi, Hirotsugu; Hirao, Masahiko

    2009-07-01

    A nonlinear acoustic measurement is studied for creep damage evaluation. An electromagnetic acoustic transducer (EMAT) magnetostrictively couples to a surface-shear-wave resonance along the circumference of a cylindrical specimen during the creep of Cr-Mo-V steels. The excitation of the EMAT at half of the resonance frequency caused a standing wave to contain only the second-harmonic component, which was received by the same EMAT for determining the second-harmonic amplitude. This measured surface-wave nonlinearity showed a peak at 30% and a minimum at 50% of the total life. We interpreted these phenomena in terms of dislocation mobility and restructuring, with support from scanning electron microscope (SEM) and transmission electron microscope (TEM) observations. This noncontact resonance-EMAT measurement can monitor the evolution of surface-shear-wave nonlinearity throughout creep life and has a potential to assess damage advance and predict the creep life of metals.

  16. Electromagnetic Acoustic Resonance to Assess Creep Damage in Cr-Mo-V Steel

    NASA Astrophysics Data System (ADS)

    Ohtani, Toshihiro; Ogi, Hirotsugu; Hirao, Masahiko

    2006-05-01

    Electromagnetic acoustic resonance (EMAR) is a contactless resonance method using an electromagnetic acoustic transducer (EMAT). In this study, EMAR was applied to detect the creep damage process in Cr-Mo-V steel, which is an important structural material for thermal energy plants. The material was exposed to temperatures up to 923 K at various stresses. Two types of EMAT were used: bulk-wave EMAT for plate samples and axial-shear-wave EMAT for cylindrical samples. We measured ultrasonic attenuation in the frequency range between 1 and 7 MHz as creep progressed. Attenuation coefficient exhibits a much larger sensitivity to damage accumulation than velocity. It shows a maximum peak at approximately 30% and a minimum peak at 50% of the creep life, independent of the applied stress and the type of EMAT used. EMAR has the potential for assessing damage progress and for predicting the creep life of metals.

  17. Optimizing the corrosion fatigue properties of Co-Cr-Mo Hip joints

    NASA Astrophysics Data System (ADS)

    Tensi, Hans M.; Hooputra, Hariaokto; Weinfurtner, Wolfgang; Mayr, Hubert

    1995-01-01

    Because of their affordability and their adaptability to different designs, cast Co-Cr-Mo alloys are the materials most used for hip joint endoprostheses. These alloys combine excellent biocompatibility with a high corrosion resistance. Most hip joint endoprostheses are manufactured by conventional casting. The microstructural defects caused by this casting method lead to premature fractures. Aseptic loosening of endoprostheses also contributes to fracture. This article shows that using unidirectional solidification prolongs the mean value of fatigue life by at least six times over conventional casting; comparing the lowest values, the fatigue life is more than ten times higher. The comparison is made for two different kinds of solidified tension-compression specimens without any heat treatment to study only the influence of the solidification process. It should also be noted, however, that heat treatment adapted to microstructural parameters can elevate fatigue life.

  18. Anisotropic elastic properties of MB (M = Cr, Mo, W) monoborides: a first-principles investigation

    NASA Astrophysics Data System (ADS)

    Li, Run-Yue; Duan, Yong-Hua

    2016-04-01

    First principles calculations were performed to systematically investigate structure properties, phase stability and mechanical properties of MB (M = Cr, Mo, W) monoborides in orthorhombic and tetragonal structures. The results of equilibrium structures are in good agreement with other available theoretical and experimental data. The elastic properties, including bulk modulus B, shear modulus G, Young's modulus E and Poisson's ratio ν were calculated by the Voigt-Reuss-Hill approximation. All considered monoborides are mechanically stable. The results of elastic anisotropies show that elastic anisotropy of orthorhombic structure is larger than that of tetragonal structure. Moreover, the minimum thermal conductivities were also estimated using the Cahill's model, and the results indicate that the minimum thermal conductivities show a dependence on directions.

  19. Characterization of Wear Particles Generated from CoCrMo Alloy under Sliding Wear Conditions

    PubMed Central

    Pourzal, R.; Catelas, I.; Theissmann, R.; Kaddick, C.; Fischer, A.

    2011-01-01

    Biological effects of wear products (particles and metal ions) generated by metal-on-metal (MoM) hip replacements made of CoCrMo alloy remain a major cause of concern. Periprosthetic osteolysis, potential hypersensitivity response and pseudotumour formation are possible reactions that can lead to early revisions. To accurately analyse the biological response to wear particles from MoM implants, the exact nature of these particles needs to be characterized. Most previous studies used energy-dispersive X-ray spectroscopy (EDS) analysis for characterization. The present study used energy filtered transmission electron microscopy (TEM) and electron diffraction pattern analysis to allow for a more precise determination of the chemical composition and to gain knowledge of the crystalline structure of the wear particles. Particles were retrieved from two different test rigs: a reciprocating sliding wear tribometer (CoCrMo cylinder vs. bar) and a hip simulator according to ISO 14242-1 (CoCrMo head vs. CoCrMo cup). All tests were conducted in bovine serum. Particles were retrieved from the test medium using a previously published enzymatic digestion protocol. Particles isolated from tribometer samples had a size of 100 – 500 nm. Diffraction pattern analysis clearly revealed the lattice structure of strain induced hcp ε-martensite. Hip simulator samples revealed numerous particles of 15 – 30 nm and 30 – 80 nm size. Most of the larger particles appeared to be only partially oxidized and exhibited cobalt locally. The smallest particles were Cr2O3 with no trace of cobalt. It optically appeared that these Cr2O3 particles were flaking off the surface of larger particles that depicted a very high intensity of oxygen, as well as chromium, and only background noise of cobalt. The particle size difference between the two test rigs is likely related to the conditions of the two tribosystems, in particular the difference in the sample geometry and in the type of sliding

  20. Creep-Induced Microstructural Changes and Acoustic Characterization in a Cr-Mo-V Steel

    NASA Astrophysics Data System (ADS)

    Ohtani, Toshihiro; Yin, Fuxing; Kamada, Yasuhiro

    2008-05-01

    We studied the evolution of microstructure in a Cr-Mo-V steel (JIS-SNB16) during creep by monitoring ultrasonic attenuation. After obtaining a series of creep samples with various strains under a tensile stress of 25 MPa at 923 K, we removed small samples from the creep samples and measured free vibration resonance frequencies and attenuation coefficients with electromagnetic acoustic resonance (EMAR). EMAR is a combination of the resonant acoustic technique with a non-contact electromagnetic acoustic transducer (EMAT). The attenuation measurement is inherently free from any energy loss, resulting in pure attenuation in a metal sample. Furthermore, we observed the evolution of microstructure with electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). The result from the small samples shows the same trend as our previous result from larger sample. We propose a non-destructive method using EMAR to evaluate creep damage in small specimens sampled from structural metals in-service.

  1. Fatigue strength of Co-Cr-Mo alloy clasps prepared by selective laser melting.

    PubMed

    Kajima, Yuka; Takaichi, Atsushi; Nakamoto, Takayuki; Kimura, Takahiro; Yogo, Yoshiaki; Ashida, Maki; Doi, Hisashi; Nomura, Naoyuki; Takahashi, Hidekazu; Hanawa, Takao; Wakabayashi, Noriyuki

    2016-06-01

    We aimed to investigate the fatigue strength of Co-Cr-Mo clasps for removable partial dentures prepared by selective laser melting (SLM). The Co-Cr-Mo alloy specimens for tensile tests (dumbbell specimens) and fatigue tests (clasp specimens) were prepared by SLM with varying angles between the building and longitudinal directions (i.e., 0° (TL0, FL0), 45° (TL45, FL45), and 90° (TL90, FL90)). The clasp specimens were subjected to cyclic deformations of 0.25mm and 0.50mm for 10(6) cycles. The SLM specimens showed no obvious mechanical anisotropy in tensile tests and exhibited significantly higher yield strength and ultimate tensile strength than the cast specimens under all conditions. In contrast, a high degree of anisotropy in fatigue performance associated with the build orientation was found. For specimens under the 0.50mm deflection, FL90 exhibited significantly longer fatigue life (205,418 cycles) than the cast specimens (112,770 cycles). In contrast, the fatigue lives of FL0 (28,484 cycles) and FL45 (43,465 cycles) were significantly shorter. The surface roughnesses of FL0 and FL45 were considerably higher than those of the cast specimens, whereas there were no significant differences between FL90 and the cast specimens. Electron backscatter diffraction (EBSD) analysis indicated the grains of FL0 showed preferential close to <001> orientation of the γ phase along the normal direction to the fracture surface. In contrast, the FL45 and FL90 grains showed no significant preferential orientation. Fatigue strength may therefore be affected by a number of factors, including surface roughness and crystal orientation. The SLM process is a promising candidate for preparing tough removable partial denture frameworks, as long as the appropriate build direction is adopted. PMID:26974490

  2. New Insights into Hard Phases of CoCrMo Metal-on-Metal Hip Replacements

    PubMed Central

    Liao, Y.; Pourzal, R.; Stemmer, P.; Wimmer, M.A.; Jacobs, J.J.; Fischer, A.; Marks, L. D.

    2012-01-01

    The microstructural and mechanical properties of the hard phases in CoCrMo prosthetic alloys in both cast and wrought conditions were examined using transmission electron microscopy and nanoindentation. Besides the known carbides of M23C6-type (M=Cr, Mo, Co) and M6C-type which are formed by either eutectic solidification or precipitation, a new mixed-phase hard constituent has been found in the cast alloys, which is composed of ~100 nm fine grains. The nanosized grains were identified to be mostly of M23C6 type using nano-beam precession electron diffraction, and the chemical composition varied from grain to grain being either Cr- or Co-rich. In contrast, the carbides within the wrought alloy having the same M23C6 structure were homogeneous, which can be attributed to the repeated heating and deformation steps. Nanoindentation measurements showed that the hardness of the hard phase mixture in the cast specimen was ~15.7 GPa, while the M23C6 carbides in the wrought alloy were twice as hard (~30.7 GPa). The origin of the nanostructured hard phase mixture was found to be related to slow cooling during casting. Mixed hard phases were produced at a cooling rate of 0.2 °C/s, whereas single phase carbides were formed at a cooling rate of 50 °C/s. This is consistent with sluggish kinetics and rationalizes different and partly conflicting microstructural results in the literature, and could be a source of variations in the performance of prosthetic devices in-vivo. PMID:22659365

  3. Microstructural characterization of as-cast biocompatible Co-Cr-Mo alloys

    SciTech Connect

    Giacchi, J.V.; Morando, C.N.; Fornaro, O.; Palacio, H.A.

    2011-01-15

    The microstructure of a cobalt-base alloy (Co-Cr-Mo) obtained by the investment casting process was studied. This alloy complies with the ASTM F75 standard and is widely used in the manufacturing of orthopedic implants because of its high strength, good corrosion resistance and excellent biocompatibility properties. This work focuses on the resulting microstructures arising from samples poured under industrial environment conditions, of three different Co-Cr-Mo alloys. For this purpose, we used: 1) an alloy built up from commercial purity constituents, 2) a remelted alloy and 3) a certified alloy for comparison. The characterization of the samples was achieved by using optical microscopy (OM) with a colorant etchant to identify the present phases and scanning electron microscopy (SE-SEM) and energy dispersion spectrometry (EDS) techniques for a better identification. In general the as-cast microstructure is a Co-fcc dendritic matrix with the presence of a secondary phase, such as the M{sub 23}C{sub 6} carbides precipitated at grain boundaries and interdendritic zones. These precipitates are the main strengthening mechanism in this type of alloys. Other minority phases were also reported and their presence could be linked to the cooling rate and the manufacturing process variables and environment. - Research Highlights: {yields}The solidification microstructure of an ASTM-F75 type alloy were studied. {yields}The alloys were poured under an industrial environment. {yields}Carbides and sigma phase identified by color metallography and scanning microscopy (SEM and EDS). {yields}Two carbide morphologies were detected 'blocky type' and 'pearlite type'. {yields}Minority phases were also detected.

  4. Plastic Deformation Behavior and Processing Maps of 35CrMo Steel

    NASA Astrophysics Data System (ADS)

    Xiao, Zheng-bing; Huang, Yuan-chun; Liu, Yu

    2016-03-01

    Hot deformation behavior of 35CrMo steel was investigated by compression tests in the temperature range of 850 to 1150 °C and strain rate range of 0.01 to 20 s-1 on a Gleeble-3810 thermal simulator. According to processing maps constructed based on the experimental data and using the principle of dynamic materials modeling (DMM), when the strain is 0.8, three safe regions with comparatively high efficiency of power dissipation were identified: (850 to 920) °C/(0.01 to 0.02) s-1, (850 to 900) °C/(10 to 20) s-1, and (1050 to 1150) °C/(0.01 to 1) s-1. And the domain of (920 to 1150) °C/(2.7 to 20) s-1 is within the instability range, whose efficiency of power dissipation is around 0.05. The deformed optical microstructure indicated that the combination of low deformation temperature (850 °C) and a relatively high strain rate (20 s-1) resulted in the smallest dynamic recrystallized grains, but coarser grains were obtained when a much higher strain rate was employed (50 s-1). A lower strain rate or a higher temperature will accelerate the growth of grains, and both high temperature and high strain rate can cause microcracks in the deformed steel. Integration of the processing map into the optical microstructure identified the region of (850 to 900) °C/(10 to 20) s-1 as the ideal condition for the hot deformation of 35CrMo steel.

  5. Effect of gas nitriding on CO2 corrosion for 35CrMo steel after surface nanocrystallization.

    PubMed

    Wang, Bingying; Zhou, Shengnan; Wang, Jingjing; Zhao, Bin

    2014-10-01

    This paper studies the influence of ultrasonic surface rolling procession (USRP) and gas nitriding on CO2 corrosion for 35CrMo steel. The microstructure of the nanocrystallized surface caused by USRP and the nitrided layer were studied by means of HRTEM and optical microscope, respectively. High temperature high pressure autoclave was adopted to study the CO2 corrosion behavior of 35CrMo steel. The characteristics of CO2 corrosion scales on 35CrMo steel were investigated by the SEM, EDS and XRD techniques. The experimental results show that after USRP about 250 μm rheological layer forms on the metal surface, and the average grain size is 25 nm. USRP thicken the nitrided layer, 10 hours' gas nitriding at 550 degrees C lower the corrosion rate while the combine of gas nitriding and USRP enhances the corrosion resistance furthest; and the surface nanocrystallization increases the content of Cr and changes the corrosion product film from FeCO3 to FeCO3 and Cr2O3, and from loose crystal structure to amorphous flocculent structure. The corrosion resistance of 35CrMo has been improved significantly by USRP and gas nitriding. PMID:25942927

  6. Notch-Fatigue Properties of Advanced TRIP-Aided Bainitic Ferrite Steels

    NASA Astrophysics Data System (ADS)

    Yoshikawa, Nobuo; Kobayashi, Junya; Sugimoto, Koh-ichi

    2012-11-01

    To develop a transformation-induced plasticity (TRIP)-aided bainitic ferrite steel (TBF steel) with high hardenability for a common rail of the next generation diesel engine, 0.2 pct C-1.5 pct Si-1.5 pct Mn-0.05 pct Nb TBF steels with different contents of Cr, Mo, and Ni were produced. The notch-fatigue strength of the TBF steels was investigated and was related to the microstructural and retained austenite characteristics. If Cr, Mo, and/or Ni were added to the base steel, then the steels achieved extremely higher notch-fatigue limits and lower notch sensitivity than base TBF steel and the conventional structural steels. This was mainly associated with (1) carbide-free and fine bainitic ferrite lath structure matrix without proeutectoid ferrite, (2) a large amount of fine metastable retained austenite, and (3) blocky martensite phase including retained austenite, which may suppress a fatigue crack initiation and propagation.

  7. Metallurgical causes for the occurrence of creep damage in longitudinally seam-welded Cr-Mo high-energy piping

    NASA Astrophysics Data System (ADS)

    Zhou, Gang

    A continuous occurrence of catastrophic failures, leaks and cracks of the Cr-Mo steam piping has created widespread utility concern for the integrity and serviceability of the seam-welded piping systems in power plants across USA. Cr-Mo steels are the materials widely used for elevated temperature service in fossil-fired generating stations. A large percentage of the power plant units with the Cr-Mo seam-welded steam piping have been in operation for a long duration such that the critical components of the units have been employed beyond the design life (30 or 40 years). This percentage will increase even more significantly in the near future. There is a strong desire to extend and thus there is a need to assess the remaining life of these units. Thus, understanding of the metallurgical causes for the failures and damage in the Cr-Mo seam-welded piping plays a major role in estimating possible life-extension and decision making on whether to operate, repair or replace. In this study, an optical metallographic method and a Cryo-Crack fractographic method have been developed for characterization and quantification of the damage in seam-welded steam piping. More than 500 metallographic assessments, from more than 25 power plants, have been accomplished using the optical metallographic method, and more than 200 fractographic specimens from 10 power plants have been evaluated using the "Cryo-Crack" fractographic technique. For comparison, "virgin" SA welds were fabricated using the Mohave welding procedure with re-N&T Mohave base metal with both "acid" and "basic" fluxes. The damage mechanism, damage distribution pattern, damage classification, correlation of the damage with the microstructural features of these SA welds and the impurity segregation patterns have been determined. A physical model for cavitation (leading to failure) in Cr-Mo SA weld metals and evaluation methodologies for high energy piping are proposed.

  8. R&D of low activation ferritic steels for fusion in japanese universities*1

    NASA Astrophysics Data System (ADS)

    Kohyama, Akira; Kohno, Yutaka; Asakura, Kentaro; Kayano, Hideo

    1994-09-01

    Following the brief review of the R&D of low activation ferritic steels in Japanese universities, the status of 9Cr-2W type ferritic steels development is presented. The main emphasis is on mechanical property changes by fast neutron irradiation in FFTF. Bend test, tensile test, CVN test and in-reactor creep results are provided including some data about low activation ferritic steels with Cr variation from 2.25 to 12%. The 9Cr-2W ferritic steel, denoted as JLF-1, showed excellent mechanical properties under fast neutron irradiation as high as 60 dpa. As potential materials for DEMO and beyond, innovative oxide dispersion strengthened (ODS) quasi-amorphous low activation ferritic steels are introduced. The baseline properties, microstructural evolution under ion irradiation and the recent progress of new processes are provided.

  9. Nano-cluster stability following neutron irradiation in MA957 oxide dispersion strengthened material

    NASA Astrophysics Data System (ADS)

    Ribis, J.; Lozano-Perez, S.

    2014-01-01

    ODS steels are promising materials for Sodium cooled Fast Reactors since their fine distribution of nano-clusters confers excellent mechanical properties. However, the nano-feature stability needs to be assessed under neutron irradiation. Before irradiation, the characterizations show that nano-particles are finely distributed within the ferritic matrix and are identified to have a pyrochlore type structure. After irradiation of the MA957 alloy in the Phenix French reactor at 412 °C up to 50 dpa and 430 °C up to 75 dpa, transmission electron microscopy characterization reveals a very slight density fall but no distinguishable difference in nano-features size before and after irradiation. In addition, after both irradiations, the nano-oxides are still (Y, Ti, O) compounds with orientation relationship with the matrix. A multislice simulation of high resolution images suggests that nano-particles still have a fcc pyrochlore type structure after irradiation. A possible change of lattice parameter seems to be highlighted, possibly due to disordering by cascade effect.

  10. Quantitative characterization of microstructural defects in up to 32 dpa neutron irradiated EUROFER97

    NASA Astrophysics Data System (ADS)

    Weiß, Oliver J.; Gaganidze, Ermile; Aktaa, Jarir

    2012-07-01

    The microstructure of the neutron-irradiated reduced activation ferritic/martensitic (RAFM) steel EUROFER97 was evaluated by transmission electron microscopy (TEM). Emphasis was put on analyzing the influence of the irradiation dose on the evolution of size and density of microstructural defects like dislocation loops and voids at low irradiation temperatures of 330-340 °C. To study the dose dependence, samples irradiated to 15 and 32 dpa were analyzed. The weak-beam dark-field (WBDF) technique was applied to analyze dislocation loops and small defect clusters using different diffraction conditions. The average densities and sizes of the defects increase slightly from 1.4 × 1022 m-3 and 3.4 nm at 15 dpa to 1.7 × 1022 m-3 and 4.8 nm at 32 dpa. Through-focus series also revealed the presence of small voids in the material, but with a density at least one order of magnitude lower than that of the dislocation loops. In order to correlate the irradiation induced changes in the microstructure to the changes in the mechanical properties, the obtained quantitative data was used to estimate dose-dependent hardening with the dispersed barrier hardening model. The estimation of hardening by using recent literature results on the loop obstacle strength shows, that alone the defects visible in the TEM are not sufficient to explain the hardening quantified in the post-irradiation tensile tests.

  11. Test of radiation hardness of CMOS transistors under neutron irradiation

    SciTech Connect

    Sadrozinski, H.F.W.; Rowe, W.A.; Seiden, A.; Spencer, E.; Hoffman, C.M.; Holtkamp, D.; Kinnison, W.W.; Sommer, W.F. Jr.; Ziock, H.J.

    1989-01-01

    We have tested 2 micron CMOS test structures from various foundries in the LAMPF Beam stop for radiation damage under prolongued neutron irradiation. The fluxes employed covered the region expected to be encountered at the SSC and led to fluences of up to 10/sup 14/ neutrons/cm/sup 2/ in about 500 hrs of running. We show that test structures which have been measured to survive ionizing radiation of the order MRad also survive these high neutron fluences. 5 refs., 4 figs.

  12. TiO2-coated CoCrMo: improving the osteogenic differentiation and adhesion of mesenchymal stem cells in vitro.

    PubMed

    Logan, Niall; Sherif, Anas; Cross, Alison J; Collins, Simon N; Traynor, Alison; Bozec, Laurent; Parkin, Ivan P; Brett, Peter

    2015-03-01

    The current gold standard material for orthopedic applications is titanium (Ti), however, other materials such as cobalt-chromium-molybdenum (CoCrMo) are often preferred due to their wear resistance and mechanical strength. This study investigates if the bioactivity of CoCrMo can be enhanced by coating the surface with titanium oxide (TiO2 ) by atmospheric pressure chemical vapor deposition (CVD), thereby replicating the surface oxide layer found on Ti. CoCrMo, TiO2-coated CoCrMo (CCMT) and Ti substrates were used for this study. Cellular f-actin distribution was shown to be noticeably different between cells on CCMT and CoCrMo after 24 h in osteogenic culture, with cells on CCMT exhibiting greater spread with developed protrusions. Osteogenic differentiation was shown to be enhanced on CCMT compared to CoCrMo, with increased calcium ion content per cell (p < 0.05), greater hydroxyapatite nodule formation (p < 0.05) and reduced type I collagen deposition per cell (p < 0.05). The expression of the focal adhesion protein vinculin was shown to be marginally greater on CCMT compared to CoCrMo, whereas AFM results indicated that CCMT required more force to remove a single cell from the substrate surface compared to CoCrMo (p < 0.0001). These data suggest that CVD TiO2 coatings may have the potential to increase the biocompatibility of CoCrMo implantable devices. PMID:25045159

  13. Dynamic Strain Aging in New Generation Cr-Mo-V Steel for Reactor Pressure Vessel Applications

    NASA Astrophysics Data System (ADS)

    Gupta, C.; Chakravartty, J. K.; Banerjee, S.

    2010-12-01

    A new generation nuclear reactor pressure vessel steel (CrMoV type) having compositional similarities with thick section 3Cr-Mo class of low alloy steels and adapted for nuclear applications was investigated for various manifestations of dynamic strain aging (DSA) using uniaxial tests. The steel investigated herein has undergone quenched and tempered treatment such that a tempered bainite microstructure with Cr-rich carbides was formed. The scope of the uniaxial experiments included tensile tests over a temperature range of 298 K to 873 K (25 °C to 600 °C) at two strain rates (10-3 and 10-4 s-1), as well as suitably designed transient strain rate change tests. The flow behavior displayed serrated flow, negative strain rate sensitivity, plateau behavior of yield, negative temperature ( T), and strain rate left( {dot{\\varepsilon }} right) dependence of flow stress over the temperature range of 523 K to 673 K (250 °C to 400 °C) and strain rate range of 5 × 10-3 s-1 to 3 × 10-6 s-1, respectively. While these trends attested to the presence of DSA, a lack of work hardening and near negligible impairment of ductility point to the fact that manifestations of embrittling features of DSA were significantly enervated in the new generation pressure vessel steel. In order to provide a mechanistic understanding of these unique combinations of manifestations of DSA in the steel, a new approach for evaluation of responsible solutes from strain rate change tests was adopted. From these experiments and calculation of activation energy by application of vacancy-based models, the solutes responsible for DSA were identified as carbon/nitrogen. The lack of embrittling features of DSA in the steel was rationalized as being due to the beneficial effects arising from the presence of dynamic recovery effects, presence of alloy carbides in the tempered bainitic structure, and formation of solute clusters, all of which hinder the possibilities for strong aging of dislocations.

  14. Microstructural evolution during solution treatment of Co-Cr-Mo-C biocompatible alloys

    SciTech Connect

    Giacchi, J.V.; Fornaro, O.; Palacio, H.

    2012-06-15

    Three different Co-Cr-Mo-C alloys conforming to ASTM F75 standard were poured in an industrial environment and subjected to a conventional solution treatment at 1225 Degree-Sign C for several time intervals. The microstructural changes and transformations were studied in each case in order to evaluate the way in which treatment time influences the secondary phase fraction and clarify the microstructural changes that could occur. To assess how treatment time affects microstructure, optical microscopy and image analyzer software, scanning electron microscopy and energy dispersion spectrometry analysis were employed. The main phases detected in the as-cast state were: {sigma}-phase, M{sub 6}C, and M{sub 23}C{sub 6} carbides. The latter presented two different morphologies, blocky type and lamellar type. Despite being considered the most detrimental feature to mechanical properties, {sigma}-phase and lamellar carbides dissolution took place in the early stages of solution treatment. M{sub 23}C{sub 6} carbides featured two different behaviors. In the alloy obtained by melting an appropriate quantity of alloyed commercial materials, a decrease in size, spheroidization and transformation into M{sub 6}C carbides were simultaneously observed. In the commercial ASTM F75 alloy, in turn, despite being the same phase, only a marked decrease in precipitates size was noticed. These different behaviors could be ascribed to the initial presence of other phases in the alloy obtained from alloyed materials, such as {sigma}-phase and 'pearlitic' carbides, or to the initial precipitate size which was much larger in the first than in the commercial ASTM F75 alloy studied. M{sub 6}C carbides dissolved directly in the matrix as they could not be detected in samples solution-treated for 15 min. - Highlights: Black-Right-Pointing-Pointer Three different Co-Cr-Mo alloys were poured under an industrial environment. Black-Right-Pointing-Pointer Transformation of existing phases followed during

  15. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    DOE PAGESBeta

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019more » n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

  16. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    SciTech Connect

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  17. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    NASA Astrophysics Data System (ADS)

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-11-01

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  18. Defect microstructures in neutron-irradiated copper and stainless steel

    SciTech Connect

    Zinkle, S.J.; Sindelar, R.L.

    1987-09-01

    The defect microstructures of copper and type 304L austenitic stainless steel have been examined following neutron irradiation under widely different conditions. Less than 0.2% of the defect clusters in steel irradiated at 120/sup 0/C with moderated fission neutrons were resolvable as stacking fault tetrahedra (SFT). The fraction of defect clusters identified as SFT in copper varied from approx.10% for a low-dose 14-MeV neutron irradiation at 25/sup 0/C to approx.50% for copper irradiated to 1.3 dpa in a moderated fission spectrum at 182/sup 0/C. The mean cluster size in copper was about 2.6 nm for both cases, despite the large differences in irradiation conditions. The mean defect cluster size in the irradiated steel was about 1.8 nm. The absence of SFT in stainless steel may be due to the generation of 35 appm He during the irradiation, which caused the vacancies to form helium-filled cavities instead of SFT. 20 refs.

  19. Impurities effect on the swelling of neutron irradiated beryllium

    SciTech Connect

    Donne, M.D.; Scaffidi-Argentina, F.

    1995-09-01

    An important factor controlling the swelling behaviour of fast neutron irradiated beryllium is the impurity content which can strongly affect both the surface tension and the creep strength of this material. Being the volume swelling of the old beryllium (early sixties) systematically higher than that of the more modem one (end of the seventies), a sensitivity analysis with the aid of the computer code ANFIBE (ANalysis of Fusion Irradiated BEryllium) to investigate the effect of these material properties on the swelling behaviour of neutron irradiated beryllium has been performed. Two sets of experimental data have been selected: the first one named Western refers to quite recently produced Western beryllium, whilst the second one, named Russian refers to relatively old (early sixties) Russian beryllium containing a higher impurity rate than the Western one. The results obtained with the ANFIBE Code were assessed by comparison with experimental data and the used material properties were compared with the data available in the literature. Good agreement between calculated and measured values has been found.

  20. Positron annihilation in neutron irradiated iron-based materials

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Almazouzi, A.

    2011-01-01

    The hardening and embrittlement of reactor pressure vessel steels is of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, like vacancies, interstitials, solutes and their clusters. But the reason for the embrittlement of the material is not yet totally known. The real nature of the irradiation damage should thus be examined as well as its evolution in time. Positron annihilation spectroscopy has been shown to be a powerful method for analyzing some of these defects. In fact, both vacancy type clusters and precipitates can be visualized by positrons. Recently, at SCK·CEN, a new setup has been constructed, calibrated and optimized to measure the coincidence Doppler broadening and lifetime of neutron irradiated materials. To be able to compare the results obtained by the positron studies, with those of other techniques (such as transmission electron microscopy, atom probe tomography and small angle neutron scattering), quantitative estimations of the size and density of the annihilation sites are needed. Using the approach proposed by Vehanen et al., an attempt is made to calculate the needed quantities in Fe and Fe-Cu binary alloys that were neutron irradiated to different doses. The results obtained are discussed highlighting the difficulties in defining the annihilation centres even in these simple model alloys, in spite of using both lifetime and Doppler broadening measurements in the same samples.

  1. Structural and electronic properties of XSi{sub 2} (X = Cr, Mo, and W)

    SciTech Connect

    Shugani, Mani; Aynyas, Mahendra; Sanyal, S. P.

    2015-07-15

    The structural and electronic properties of metal silicides XSi{sub 2} (X = Cr, Mo, and W), which crystallize in tetragonal structure, are investigated systematically using the first-principle density functional theory. The total energies are computed as a function of volume and fitted to the Birch equation of state. The ground-state properties such as equilibrium lattice constants a{sub 0} and c{sub 0}, bulk modulus B, its pressure derivative B, B′, and the density of states at the Fermi level, N(E{sub F}), are calculated and compared with other experimental and theoretical results, showing good agreement. The calculated band structure indicates that XSi{sub 2} compounds are semimetallic in nature. From the present study, we predict the structural and electronic properties of CrSi{sub 2} in the tetragonal phase and indicate that CrSi{sub 2} is energetically more stable than MoSi{sub 2} and WSi{sub 2}. Analyzing the bonding properties of the three metal silicides, we observe that WSi{sub 2} has a strong covalent bonding due to W 5d electrons.

  2. Improvement of Ni-Cr-Mo coating performance by laser cladding combined re-melting

    NASA Astrophysics Data System (ADS)

    Wang, Qin-Ying; Bai, Shu-Lin; Zhang, Yang-Fei; Liu, Zong-De

    2014-07-01

    Although being an efficient technique to produce metallic alloy coating, laser cladding may leave original unmelted particles in the coating. Further treatment is thus necessary to improve the coating quality, and laser re-melting therefore becomes a potential method. In this study, Ni-Cr-Mo alloy coatings were prepared on Q235 steel substrate by laser cladding (coating N1) and then re-melted by laser (coating N2) with the same technic parameters. The initial defect evolution and its effect on hardness and corrosion resistance of coatings were studied. The results show that there are fewer and smaller defects in coating N2 than in coating N1, which is ascribbed to the disappearance and partial melting of Cr/Cr2O3 particles. The nearly unchanged hardness of coatings N1 and N2 is justified by both Vickers tests and nanoindentation combined theoretical calculation. Coating N2 with higher positive corrosion potential and lower corrosion current density exhibits better corrosion resistance than coating N1. Above results prove that laser re-melting can refine the microstructure and improve corrosion resistance of coatings to some degree.

  3. The Kinetics of Metadynamic Recrystallization in a Ni-Cr-Mo-Based Superalloy Hastelloy C-276

    NASA Astrophysics Data System (ADS)

    Zhang, Chi; Zhang, Liwen; Shen, Wenfei; Liu, Cuiru; Xia, Yingnan

    2016-02-01

    The metadynamic recrystallization (MDRX) behavior of a typical Ni-Cr-Mo-based superalloy Hastelloy C-276 was investigated using two-stage isothermal compression tests on a Gleeble thermal-mechanical simulator in the temperature range of 1050-1200 °C, the strain rate range of 0.1-5.0 s-1, the strains of 0.32, 0.45, and 0.6 at the first stage of compression, and the interval times of 0.5-30 s. The results show that the microstructure and the stress-strain relation of the studied superalloy vary during the interruption period due to the occurrence of MDRX. The MDRX softening fraction and recrystallized grain size increase rapidly with the increasing of interval time, deformation temperature, and strain rate. The effect of strain at the first stage of compression on MDRX is less pronounced. The kinetics of MDRX softening was established based on the flow stress curves, and the apparent activation energy of MDRX of Hastelloy C-276 is evaluated as 241 kJ/mol.

  4. Rotational Conformers of Group VI Metal (Cr, Mo, and w) Bis(mesitylene) Sandwich Complexes

    NASA Astrophysics Data System (ADS)

    Kumari, Sudesh; Yang, Dong-Sheng

    2010-06-01

    Group VI metal bis(mesitylene) sandwich complexes were produced by interactions between laser-vaporized metal atoms and mesitylene vapor in pulsed molecular beams, identified by photoionization time-of-flight mass spectrometry, and studied by pulsed-field-ionization zero-electron-kinetic-energy spectroscopy and density functional theory calculations. Although transition metal bis(arene) sandwiches may adopt eclipsed and staggered conformations, the group VI metal bis(mesitylene) complexes were determined to be in the eclipsed form. In this configuration, two rotational conformers, with methyl group dihedral angles of 0° and 60°, were identified for each complex. The adiabatic ionization energies of the 0° and 60° rotamers were measured to be 40557/40359, 42138/41697, and 41452/41000 cm-1 for the Cr, Mo, and W complexes, with the uncertainty of ˜{5 cm-1}. The ground electronic states of the 0°(D3h)/60° (D3d) rotamers are 1A'1/ 1A1g in the neutral form and ^2A'1/2A1g in the ionized form.

  5. Microstructure and phase evolution in laser clad chromium carbide-NiCrMoNb

    NASA Astrophysics Data System (ADS)

    Venkatesh, L.; Samajdar, I.; Tak, Manish; Doherty, Roger D.; Gundakaram, Ravi C.; Prasad, K. Satya; Joshi, S. V.

    2015-12-01

    Microstructural development in laser clad layers of Chromium carbide (CrxCy)-NiCrMoNb on SA 516 steel has been investigated. Although the starting powder contained both Cr3C2 and Cr7C3, the clad layers showed only the presence of Cr7C3. Microtexture measurements by electron back scattered diffraction (EBSD) revealed primary dendritic Cr7C3 with Ni rich FCC metallic phase being present in the interdendritic spaces. Further annealing of the laser clad layers and furnace melting of the starting powder confirmed that Cr7C3 is the primary as well as stable carbide phase in this multi component system. Increase in laser power and scanning speed progressively reduced carbide content in the laser clad layers. Increased scanning speed, which enhances the cooling rate, also led to reduction in the secondary arm spacing (λ2) of the Cr7C3 dendrites. The clad layer hardness increased with carbide content and with decreased dendrite arm spacing.

  6. Thermal-microstructural analysis of multipass welding of Cr/Mo steels

    SciTech Connect

    Oddy, A.S.; McDill, J.M.J.; Braid, J.E.M.

    1996-12-31

    Full weave repair techniques for Cr/Mo steels without post-weld heat treatment are the subject of many research programs. Coupled thermal-microstructural analyses could preselect candidate welding parameters and reduce the cost and time required. In multipass welds, microstructural simulations require transient reaustenization, austenite decomposition for arbitrary thermal cycles including reheating and tempering. Finite element thermal analysis of a three-layer, weaved weld and the microstructural analysis of the heat-affected zone (HAZ) are described. Significant variation is found in properties governing reaustenization, austenite grain growth, austenite decomposition and hardness. Hardness measurements vary by up to {+-}30 HV on the same sample. Alloy differences within the allowable range lead to HAZ hardness variations of 30 HV in multipass welds. Predicted HAZ hardnesses of the three-layer weld were in good agreement with measurements. The final microstructure was also in good agreement with experiment. The predicted HAZ width was slightly wider than was measured. This difference is easily accounted for by the variation reported in weld parameters.

  7. Overview of a Welding Development Program for a Ni-Cr-Mo-Gd Alloy

    SciTech Connect

    W. L. Hurt; R. E. Mizia; D. E. Clark

    2007-06-01

    The National Spent Nuclear Fuel Program (NSNFP), located at the Idaho National Laboratory, coordinates and integrates management and disposal of U.S. Department of Energy-owned spent nuclear fuel. These management functions include using the DOE standardized canister for packaging, storage, treatment, transport, and long-term disposal in the Yucca Mountain Repository. Nuclear criticality must be prevented in the postulated event where a waste package is breached and water (neutron moderator) is introduced into the waste package. Criticality control will be implemented by using a new, weldable, corrosion-resistant, neutron-absorbing material to fabricate the welded structural inserts (fuel baskets) that will be placed in the standardized canister. The new alloy is based on the Ni-Cr-Mo alloy system with a gadolinium addition. Gadolinium was chosen as the neutron absorption alloying element because of its high thermal neutron absorption cross section. This paper describes a weld development program to qualify this new material for American Society of Mechanical Engineers (ASME) welding procedures, develop data to extend the present ASME Code Case (unwelded) for welded construction, and understand the weldability and microstructural factors inherent to this alloy.

  8. Investigation of mechanical properties for open cellular structure CoCrMo alloy fabricated by selective laser melting process

    NASA Astrophysics Data System (ADS)

    Azidin, A.; Taib, Z. A. M.; Harun, W. S. W.; Che Ghani, S. A.; Faisae, M. F.; Omar, M. A.; Ramli, H.

    2015-12-01

    Orthodontic implants have been a major focus through mechanical and biological performance in advance to fabricate shape of complex anatomical. Designing the part with a complex mechanism is one of the challenging process and addition to achieve the balance and desired mechanical performance brought to the right manufacture technique to fabricate. Metal additive manufacturing (MAM) is brought forward to the newest fabrication technology in this field. In this study, selective laser melting (SLM) process was utilized on a medical grade cobalt-chrome molybdenum (CoCrMo) alloy. The work has focused on mechanical properties of the CoCrMo open cellular structures samples with 60%, 70%, and 80% designed volume porosity that could potentially emulate the properties of human bone. It was observed that hardness values decreased as the soaking time increases except for bottom face. For compression test, 60% designed volume porosity demonstrated highest ultimate compressive strength compared to 70% and 80%.

  9. Towards near-permanent CoCrMo prosthesis surface by combining micro-texturing and low temperature plasma carburising.

    PubMed

    Dong, Yangchun; Svoboda, Petr; Vrbka, Martin; Kostal, David; Urban, Filip; Cizek, Jan; Roupcova, Pavla; Dong, Hanshan; Krupka, Ivan; Hartl, Martin

    2015-03-01

    An advanced surface engineering process combining micro-texture with a plasma carburising process was produced on CoCrMo femoral head, and their tribological properties were evaluated by the cutting-edge pendulum hip joint simulator coupled with thin film colorimetric interferometry. FESEM and GDOES showed that precipitation-free C S-phase with a uniform case depth of 10μm was formed across the micro-textures after duplex treatment. Hip simulator tests showed that the friction coefficient was reduced by 20% for micro-metre sized texture, and the long-term tribological property of microtexture was enhanced by the C-supersaturated crystalline microstructure formed on the surface of duplex treated CoCrMo, thereby enhancing biotribological durability significantly. In-situ colorimetric interferometry confirmed that the maximum film thickness around texture area was 530nm, indicating that the additional lubricant during sliding motion might provide exceptional bearing life. PMID:26594781

  10. Optical absorption and luminescence in neutron-irradiated, silica-based fibers

    SciTech Connect

    Cooke, D.W.; Farnum, E.H.; Clinard, F.W.

    1995-04-01

    The objectives of this work are to assess the effects of thermal annealing and photobleaching on the optical absorption of neutron-irradiated, silica fibers of the type proposed for use in ITER diagnostics, and to measure x-ray induced luminescence of unirradiated (virgin) and neutron-irradiated fibers.

  11. Electrochemical investigation of chromium oxide-coated Ti-6Al-4V and Co-Cr-Mo alloy substrates.

    PubMed

    Swaminathan, Viswanathan; Zeng, Haitong; Lawrynowicz, Daniel; Zhang, Zongtao; Gilbert, Jeremy L

    2011-08-01

    Hard coatings for articulating surfaces of total joint replacements may improve the overall wear resistance. However, any coating approach must take account of changes in corrosion behavior. This preliminary assessment analyzes the corrosion kinetics, impedance and mechanical-electrochemical stability of 100 μm thick plasma sprayed chromium oxide (Cr₂O₃) coatings on bearing surfaces in comparison to the native alloy oxide films on Co-Cr-Mo and Ti-6Al-6V. Cyclic potentiodynamic polarization, electrochemical impedance spectroscopy, and mechanical abrasion under potentiostatic conditions were performed on coated and substrate surfaces in physiological saline. SEM analysis characterized the coating morphology. The results showed that the corrosion current density values of chromium oxide coatings (0.4-1.2 μA/cm²) were of the same order of magnitude as Ti-6Al-4V alloy. Mechanical abrasion did not increase corrosion rates of chromium oxide coatings but did for uncoated Co-Cr-Mo and Ti-6Al-4V. The impedance response of chromium oxide coatings was very different than Co-Cr-Mo and Ti-6Al-4V native oxides characterized by a defected coating model. More of a frequency-independent purely resistive response was seen in mid-frequency range for the coatings (CPE(coat) : 40-280 nF/cm² (rad/s)(1-α) , α: 0.67-0.83) whereas a more capacitive character is seen for Co-Cr-Mo and Ti-6Al-4V (CPE(ox) around 20 μF/cm² (rad/s)(1-α) , α around 0.9). Pores, interparticle gaps and incomplete fusion typical for thermal spray coatings were present in these oxides which could have influenced corrosion resistance. The coating microstructure could have allowed some fluid penetration. Overall, these coatings appear to have suitable corrosion properties for wear surfaces. PMID:21648063

  12. Manufacturing of 9CrMoCoB Steel of Large Ingot with Homogeneity by ESR Process

    NASA Astrophysics Data System (ADS)

    Kim, D. S.; Lee, G. J.; Lee, M. B.; Hur, J. I.; Lee, J. W.

    2016-07-01

    In case of 9CrMoCoB (COST FB2) steel, equilibrium relation between [B]/[Si] ratio and (B2O3)/(SiO2) ratio is very important to control [Si] and [B] in optimum range. Therefore, in this work, to investigate the thermodynamic equilibrium relation between [B]/[Si] ratio and (B2O3)/(SiO2) ratio, pilot ESR experiments of 9CrMoCoB steel were carried out using the CaF2-CaO-Al2O3-SiO2-B2O3 slag system according to change of Si content in electrode and B2O3 content in the slag. Furthermore, through the test melting of the 20ton-class ESR ingot, the merits and demerits of soft arcing were investigated. From these results, it is concluded that oxygen content in the ESR ingot decrease with decreasing SiO2 content in the slag, relation function between [B]/[Si] ratio and (B2O3)/(SiO2) ratio derived by Pilot ESR test shows a good agreement as compared to the calculated line with a same slope and soft arcing makes interior and surface quality of ingot worse. With the optimized ESR conditions obtained from the present study, a 1000mm diameter (20 tons) and 2200mm diameter (120ton) 9CrMoCoB steel of the ESR ingot were successfully manufactured with good homogeneity by the ESR process.

  13. A nondestructive method for estimation of the fracture toughness of CrMoV rotor steels based on ultrasonic nonlinearity.

    PubMed

    Jeong, Hyunjo; Nahm, Seung-Hoon; Jhang, Kyung-Young; Nam, Young-Hyun

    2003-09-01

    The objective of this paper is to develop a nondestructive method for estimating the fracture toughness (K(IC)) of CrMoV steels used as the rotor material of steam turbines in power plants. To achieve this objective, a number of CrMoV steel samples were heat-treated, and the fracture appearance transition temperature (FATT) was determined as a function of aging time. Nonlinear ultrasonics was employed as the theoretical basis to explain the harmonic generation in a damaged material, and the nonlinearity parameter of the second harmonic wave was the experimental measure used to be correlated to the fracture toughness of the rotor steel. The nondestructive procedure for estimating the K(IC) consists of two steps. First, the correlations between the nonlinearity parameter and the FATT are sought. The FATT values are then used to estimate K(IC) using the K(IC) versus excess temperature (i.e., T-FATT) correlation that is available in the literature for CrMoV rotor steel. PMID:12919690

  14. Direct In Vivo Inflammatory Cell-Induced Corrosion of CoCrMo Alloy Orthopedic Implant Surfaces

    PubMed Central

    Gilbert, Jeremy L.; Sivan, Shiril; Liu, Yangping; Kocagöz, Sevi; Arnholt, Christina; Kurtz, Steven M.

    2014-01-01

    Cobalt-chromium-molybdenum alloy, used for over four decades in orthopedic implants, may corrode and release wear debris into the body during use. These degradation products may stimulate immune and inflammatory responses in vivo. We report here on evidence of direct inflammatory cell-induced corrosion of human implanted and retrieved CoCrMo implant surfaces. Corrosion morphology on CoCrMo implant surfaces, in unique and characteristic patterns, and the presence of cellular remnants and biological materials intimately entwined with the corrosion indicates direct cellular attack under the cell membrane region of adhered and/or migrating inflammatory cells. Evidence supports a Fenton-like reaction mechanism driving corrosion in which reactive oxygen species are the major driver of corrosion. Using in vitro tests, large increases in corrosion susceptibility of CoCrMo were seen (40 to 100 fold) when immersed in phosphate buffered saline solutions modified with hydrogen peroxide and HCl to represent the chemistry under inflammatory cells. This discovery raises significant new questions about the clinical consequences of such corrosion interactions, the role of patient inflammatory reactions, and the detailed mechanisms at play. PMID:24619511

  15. Effect of Withdrawal Rate and Gd on the Microstructures of Directionally Solidified NiAl-Cr(Mo) Hypereutectic Alloy

    NASA Astrophysics Data System (ADS)

    Wang, Lei; Shen, Jun; Zhang, Yun-Peng; Guo, Lan-Lan

    2016-03-01

    The microstructures of Ni-31Al-32Cr-6Mo- xGd hypereutectic alloy were investigated at the withdrawal rates of 10 μm/s, 30 μm/s, and 90 μm/s. For the Gd-free hypereutectic alloy, the Cr(Mo) primary dendrites appear at the beginning of solidification when the withdrawal rate is 10 μm/s. As the solidification proceeds, the Cr(Mo) primary dendrite is eliminated, and the fully eutectic structure can be obtained in the steady-state zone. With increasing the withdrawal rate, the Cr(Mo) primary dendrites decrease gradually, and vanish at 90 μm/s. In addition, at a moderate withdrawal rate (30 μm/s), an optimum addition of Gd content (0.1 wt.%) results in the refinement of the microstructure, including the refinement of the eutectic cells and the intercellular region. Meanwhile, the new white phase ((Al x Gd1- x )2O3) appears in the boundary of the eutectic cells when the Gd content is not less than 0.1 wt.%.

  16. Metal release and speciation of released chromium from a biomedical CoCrMo alloy into simulated physiologically relevant solutions.

    PubMed

    Hedberg, Yolanda; Odnevall Wallinder, Inger

    2014-05-01

    The objective of this study was to investigate the extent of released Co, Cr(III), Cr(VI), and Mo from a biomedical high-carbon CoCrMo alloy exposed in phosphate-buffered saline (PBS), without and with the addition of 10 µM H2 O2 (PBS + H2 O2 ), and 10 g L(-1) bovine serum albumin (PBS + BSA) for time periods up to 28 days. Comparative studies were made on AISI 316L for the longest time period. No Cr(VI) release was observed for any of the alloys in either PBS or PBS + H2 O2 at open-circuit potential (no applied potential). However, at applied potentials (0.7 V vs. Ag/AgCl), Cr was primarily released as Cr(VI). Co was preferentially released from the CoCrMo alloy at no applied potential. As a consequence, Cr was enriched in the utmost surface oxide reducing the extent of metal release over time. This passivation effect was accelerated in PBS + H2 O2 . As previously reported for 316L, BSA may also enhance metal release from CoCrMo. However, this was not possible to verify due to the precipitation of metal-protein complexes with reduced metal concentrations in solution as a consequence. This was particularly important for Co-BSA complexes after sufficient time and resulted in an underestimation of metals in solution. PMID:24155151

  17. The Tribological Difference between Biomedical Steels and CoCrMo-Alloys

    PubMed Central

    Fischer, Alfons; Weiß, Sabine; Wimmer, Markus A.

    2012-01-01

    In orthopedic surgery different self-mating metal couples are used for sliding wear applications. Despite the fact that in mechanical engineering self-mating austenitic alloys often lead to adhesion and seizure in biomedical engineering the different grades of Co-base alloys show good clinical results e.g. as hip joints. The reason stems from the fact that they generate a so-called tribomaterial during articulation, which consists of a mixture of nanometer small metallic grains and organic substances from the interfacial medium, which act as boundary lubricant. Even though stainless steels also generate such a tribomaterial they were ruled out from the beginning already in the 1950 as “inappropriate”. On the basis of materials with a clinical track record this contribution shows that the cyclic creep characteristics within the shear zone underneath the tribomaterial are another important criterion for a sufficient wear behavior. By means of sliding wear and torsional fatigue tests followed by electron microscopy it is shown, that austenitic materials generate wear particles of either nano- or of microsize. The latter are produced by crack initiation and propagation within the shear fatigue zone which is related to the formation of subsurface dislocation cells and, therefore, by the fact that a Ni-containing CrNiMo solid solution allows for wavy-slip. In contrast to this a Ni-free CrMnMo solid solution with further additions of C and N only shows planar slip. This leads to the formation of nanosize wear particles and distinctly improves the wear behavior. Still the latter does not fully achieve that of CoCrMo, which also shows solely planar-slip behavior. This explains why for metallurgical reasons the Ni-containing 316L-type of steels had to fail in such boundary lubricated sliding wear tribosystems. PMID:22498283

  18. The tribological difference between biomedical steels and CoCrMo-alloys.

    PubMed

    Fischer, Alfons; Weiss, Sabine; Wimmer, Markus A

    2012-05-01

    In orthopedic surgery, different self-mating metal couples are used for sliding wear applications. Despite the fact that in mechanical engineering, self-mating austenitic alloys often lead to adhesion and seizure in biomedical engineering, the different grades of Co-base alloys show good clinical results, e.g., as hip joints. The reason stems from the fact that they generate a so-called tribomaterial during articulation, which consists of a mixture of nanometer small metallic grains and organic substances from the interfacial medium, which act as a boundary lubricant. Even though stainless steel also generate such a tribomaterial, they were ruled out from the beginning already in the 1950s as "inappropriate". On the basis of materials with a clinical track record, this contribution shows that the cyclic creep characteristics within the shear zone underneath the tribomaterial are another important criterion for a sufficient wear behavior. By means of sliding wear and torsional fatigue tests followed by electron microscopy, it is shown that austenitic materials generate wear particles of either nano- or of microsize. The latter are produced by crack initiation and propagation within the shear fatigue zone which is related to the formation of subsurface dislocation cells and, therefore, by the fact that an Ni-containing CrNiMo solid solution allows for wavy-slip. In contrast to this, an Ni-free CrMnMo solid solution with further additions of C and N only shows planar slip. This leads to the formation of nanosize wear particles and distinctly improves the wear behavior. Still, the latter does not fully achieve that of CoCrMo, which also shows a solely planar-slip behavior. This explains why for metallurgical reasons the Ni-containing 316L-type of steels had to fail in such boundary lubricated sliding wear tribosystems. PMID:22498283

  19. The effects of sulfate reducing bacteria on stainless steel and Ni-Cr-Mo alloy weldments

    SciTech Connect

    Petersen, T.A.; Taylor, S.R.

    1995-10-01

    Previous research in this laboratory demonstrated a direct correlation between alloy composition and corrosion susceptibility of stainless steel and Ni-Cr-Mo alloy weldments exposed to lake water augmented with sulfate reducing bacteria (SRB). It was shown that lake water containing an active SRB population reduced the polarization resistance (R{sub p}) on all alloys studied including those with 9% Mo. In addition, preliminary evidence indicated that edge preparation and weld heat input were also important parameters in determining corrosion performance. This prior research, however, looked at ``doctored`` weldments in which the thermal oxide in the heat affected zone was removed. The objectives of the research presented here are to further confirm these observations using as-received welds. The materials examined (listed in increasing alloy content) are 1/4 inch thick plates of 316L, 317L, AL6XN (6% Mo), alloy 625 clad steel, alloy 625, and alloy 686. Materials were welded using the tungsten inert gas (TIG) process in an argon purged environment. In addition, 317L was welded in air to test oxide effects. All samples were prepared for welding by grinding to a V-edge, except the 625 clad steel samples which were prepared using a J-edge. Electrochemical performance of welded samples was monitored in four glass cells which could each allow exposure of 8 samples to the same environment. Two cells contained lake water inoculated with SRS, and two cells contained sterilized lake water. The open circuit potential (E{sub oc}) and R{sub p} was used to correlate corrosion susceptibility and bacterial activity with alloy composition and welding parameters.

  20. Tritium release properties of neutron-irradiated Be 12Ti

    NASA Astrophysics Data System (ADS)

    Uchida, M.; Ishitsuka, E.; Kawamura, H.

    2002-12-01

    Be 12Ti has a high melting point and good chemical stability and is a promising advanced material for the neutron multiplier of the DEMO reactor that requires temperatures higher than 600 °C in a blanket. To evaluate the tritium inventory in the breeding blanket, a tritium release experiment of neutron-irradiated Be 12Ti with a total fast fluence of about 4×10 20 n/cm 2 ( E>1 MeV) was carried out at 330, 400 and 500 °C. It was clear that tritium could be released easier than from beryllium, and the apparent diffusion coefficient in Be 12Ti was about two orders larger than that in beryllium at 600-100 °C. In addition to the good tritium release property, the swelling calculated from the density change of the specimens up to 1100 °C in this test was smaller than that of beryllium.

  1. Neutron irradiation and compatibility testing of Li 2O

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Krsul, J. R.; Laug, M. T.; Walters, L. C.; Tetenbaum, M.

    1984-05-01

    A study was made of the neutron irradiation behavior of 6Li-enriched Li 2O in EBR-II. In addition, a stress corrosion study was performed ex-reactor to test the compatibility of Li 2O with a variety of stainless steels. The irradiation tests showed that tritium and helium retention in the Li 2O (˜ 89% dense) lessened with neutron exposure, and the retentions appear to approach a steady-state after ˜ 1% 6Li burnup. The stress corrosion studies, using 316 stainless steel (Ti-modified) and a 35% Ni alloy, showed that stress does not enhance the corrosion, and that dry Li 2O is not significantly corrosive, the LiOH content producing the corrosive effects. Corrosion, in general, was not severe because a passivation in sealed capsules seemed to occur after a time which greatly reduced corrosion rates.

  2. Fast neutron irradiation for advanced tumors in the pelvis

    SciTech Connect

    Battermann, J.J.; Breur, K.

    1981-08-01

    Since the end of 1975, fast neutron irradiation has been used in the Antoni van Leeuwenhoek Hospital for the treatment of advanced tumors, which had no prospect of cure by other treatment modalities. Fifty-nine patients were irradiated in the pelvic area, 22 for inoperable bladder cancer, 25 for rectal and 12 for gynecological cancer. Treatments were given 5 times per week with a 14 MeV d + T neutron generator. Persisting complete tumor regression was achieved in 11 of 22 bladded patients, 14 of 25 rectum patients and 6 of 12 gynecological patients. Because of unfavorable beam characteristics, 15 of 59 (25%) treated patients had severe radiation-induced intestinal and skin complications.

  3. New E‧ centers in neutron-irradiated α-quartz

    NASA Astrophysics Data System (ADS)

    Mashkovtsev, R. I.; Pan, Y.

    2016-03-01

    Several E‧-type defects have been revealed in neutron-irradiated natural and synthetic α-quartz by using electron paramagnetic resonance (EPR) spectroscopy. For the known E'2 center the primary spin Hamiltonian parameter matrices g and A(29Si) (hyperfine interaction with 29Si) have been refined and provide compelling evidence for spin trapping on the long-bond Si atom. The EPR spectra of the new E'11 center demonstrate that the super-hyperfine structure arises from the interaction with 27Al, the first-ever example of Al-associated E‧ centers in crystalline quartz. The matrices g and A(29Si) of E'11 and another new center (E'12) support the forward-oriented configuration proposed for the E'α center in amorphous silica.

  4. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; Miller, M. K.; Powers, K. A.; Nanstad, R. K.

    2016-03-01

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m-2 (E > 1 MeV), and inlet temperatures of ∼289 °C (∼552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m-3, this copper level was below the solubility limit. A number density of 2 × 1022 m-3 of Ni-, Mn- Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m-3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m-3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain boundary in the low fluence

  5. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGESBeta

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  6. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    SciTech Connect

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

  7. Optical absorption and luminescence studies of fast neutron-irradiated complex oxides for jewellery applications

    NASA Astrophysics Data System (ADS)

    Mironova-Ulmane, N.; Skvortsova, V.; Popov, A. I.

    2016-07-01

    We studied the optical absorption and luminescence of agate (SiO2), topaz (Al2[SiO4](F,OH)2), beryl (Be3Al2Si6O18), and prehnite (Ca2Al(AlSi3O10)(OH)2) doped with different concentrations of transition metal ions and exposed to fast neutron irradiation. The exchange interaction between the impurity ions and the defects arising under neutron irradiation causes additional absorption as well as bands' broadening in the crystals. These experimental results allow us to suggest the method for obtaining new radiation-defect induced jewellery colors of minerals due to neutron irradiation.

  8. 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.

  9. Protecting Intestinal Epithelial Cell Number 6 against Fission Neutron Irradiation through NF-κB Signaling Pathway

    PubMed Central

    Chang, Gong-Min; Gao, Ya-Bing; Wang, Shui-Ming; Xu, Xin-Ping; Zhao, Li; Zhang, Jing; Li, Jin-Feng; Wang, Yun-Liang; Peng, Rui-Yun

    2015-01-01

    The purpose of this paper is to explore the change of NF-κB signaling pathway in intestinal epithelial cell induced by fission neutron irradiation and the influence of the PI3K/Akt pathway inhibitor LY294002. Three groups of IEC-6 cell lines were given: control group, neutron irradiation of 4Gy group, and neutron irradiation of 4Gy with LY294002 treatment group. Except the control group, the other groups were irradiated by neutron of 4Gy. LY294002 was given before 24 hours of neutron irradiation. At 6 h and 24 h after neutron irradiation, the morphologic changes, proliferation ability, apoptosis, and necrosis rates of the IEC-6 cell lines were assayed and the changes of NF-κB and PI3K/Akt pathway were detected. At 6 h and 24 h after neutron irradiation of 4Gy, the proliferation ability of the IEC-6 cells decreased and lots of apoptotic and necrotic cells were found. The injuries in LY294002 treatment and neutron irradiation group were more serious than those in control and neutron irradiation groups. The results suggest that IEC-6 cells were obviously damaged and induced serious apoptosis and necrosis by neutron irradiation of 4Gy; the NF-κB signaling pathway in IEC-6 was activated by neutron irradiation which could protect IEC-6 against injury by neutron irradiation; LY294002 could inhibit the activity of IEC-6 cells. PMID:25866755

  10. Carbide Formation and Dissolution in Biomedical Co-Cr-Mo Alloys with Different Carbon Contents during Solution Treatment

    NASA Astrophysics Data System (ADS)

    Mineta, Shingo; Namba, Shigenobu; Yoneda, Takashi; Ueda, Kyosuke; Narushima, Takayuki

    2010-08-01

    The microstructures of as-cast and heat-treated biomedical Co-Cr-Mo (ASTM F75) alloys with four different carbon contents were investigated. The as-cast alloys were solution treated at 1473 to 1548 K for 0 to 43.2 ks. The precipitates in the matrix were electrolytically extracted from the as-cast and heat-treated alloys. An M23C6 type carbide and an intermetallic σ phase (Co(Cr,Mo)) were detected as precipitates in the as-cast Co-28Cr-6Mo-0.12C alloy; an M23C6 type carbide, a σ phase, an η phase (M6C-M12C type carbide), and a π phase (M2T3X type carbide with a β-manganese structure) were detected in the as-cast Co-28Cr-6Mo-0.15C alloy; and an M23C6 type carbide and an η phase were detected in the as-cast Co-28Cr-6Mo-0.25C and Co-28Cr-6Mo-0.35C alloys. After solution treatment, complete precipitate dissolution occurred in all four alloys. Under incomplete precipitate dissolution conditions, the phase and shape of precipitates depended on the heat-treatment conditions and the carbon content in the alloys. The π phase was detected in the alloys with carbon contents of 0.15, 0.25, and 0.35 mass pct after heat treatment at high temperature such as 1548 K for a short holding time of less than 1.8 ks. The presence of the π phase in the Co-Cr-Mo alloys has been revealed in this study for the first time.

  11. Temperature Effects on the Mechanical Properties of Candidate SNS Target Container Materials after Proton and Neutron Irradiation

    SciTech Connect

    Byun, T.S.

    2001-11-09

    This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54 to 2.53 dpa. Irradiation temperatures were in the range 30 to 100 C. Tensile testing was performed at room temperature (20 C) and 164 C to study the effects of test temperature on the tensile properties. Test materials displayed significant radiation-induced hardening and loss of ductility due to irradiation. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative strain hardening. In the EC316LN stainless steel, increasing the test temperature from 20 C to 164 C decreased the strength by 13 to 18% and the ductility by 8 to 36%. The tensile data for the EC316LN stainless steel irradiated in spallation conditions were in line with the values in a database for 316 stainless steels for doses up to 1 dpa irradiated in fission reactors at temperatures below 200 C. However, extra strengthening induced by helium and hydrogen contents is evident in some specimens irradiated to above about 1 dpa. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. It was estimated that the 316 stainless steels would retain more than 1% true stains to necking at 164 C after irradiation to 5 dpa. A calculation using reduction of area (RA) measurements and stress-strain data predicted positive strain hardening during plastic instability.

  12. Strain Energy Approach for Axial and Torsional Fatigue Life Prediction in Aged NiCrMoV Steels

    NASA Astrophysics Data System (ADS)

    Song, Gee Wook; Hyun, Jung Seob; Ha, Jeong Soo

    Axial and torsional low cycle fatigue tests were performed for NiCrMoV steels serviced low-pressure turbine rotor of nuclear power plant. The results were used to evaluate multiaxial fatigue life models including Tresca, von Mises and Brown and Miller's critical plane. The fatigue life predicted by the multiaxial fatigue models didn't correspond with the experimental results in small strain range. We proposed the total strain energy density model to predict torsional fatigue life from axial fatigue data. The total strain energy density model was found to best correlate the experimental data with predictions being within a factor of 2.

  13. Effect of neutron energy and fluence on deuterium retention behaviour in neutron irradiated tungsten

    NASA Astrophysics Data System (ADS)

    Fujita, Hiroe; Yuyama, Kenta; Li, Xiaochun; Hatano, Yuji; Toyama, Takeshi; Ohta, Masayuki; Ochiai, Kentaro; Yoshida, Naoaki; Chikada, Takumi; Oya, Yasuhisa

    2016-02-01

    Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe2+ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe2+ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

  14. PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

    DOEpatents

    Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

    1959-11-10

    Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

  15. Precipitate stability in neutron-irradiated Zircaloy-4

    NASA Astrophysics Data System (ADS)

    Yang, W. J. S.

    1988-09-01

    Zircaloy-4, a zirconium-base alloy used extensively as cladding and core structural materials in water-cooled nuclear reactors, was examined by transmission electron microscopy, after neutron irradiation and postirradiation annealing. Phase instabilities found during irradiation at 561 K include the amorphous transformation and the dissolution of the intermetallic Zr(Fe,Cr) 2. The α-matrix is driven toward a single phase solid solution as the neutron fluence increases. This is evidenced by the continuous dissolution of the precipitate without precipitation of any new phase during irradiation. During postirradiation annealing at 833 K, solute Fe precipitates out particularly at the grain boundaries as Zr-Fe zeta-phase. Recrystallization of the amorphous precipitates occurs at a postirradiation annealing temperature of 1023 K. In general, the observed phenomena of amorphous transformation, precipitate dissolution, reprecipitation and recrystallization reflect the complex solute-point defect interactions in the α-matrix. The continuous solute dissolution during irradiation is expected to have a potential effect on irradiation growth, creep and corrosion properties of the alloy.

  16. The effect of neutron irradiation on silicon carbide fibers

    SciTech Connect

    Newsome, G.A.

    1997-01-01

    Nine types of SiC fiber have been exposed to neutron radiation in the Advanced Test Reactor at 250 C for various lengths of time ranging from 83 to 128 days. The effects of these exposures have been initially determined using scanning electron microscopy. The fibers tested were Nicalon{trademark} CG, Tyranno, Hi-Nicalon{trademark}, Dow Corning SiC, Carborundum SiC, Textron SCS-6, polymethysilane (PMS) derived SiC from the University of Michigan, and two types of MER SiC fiber. This covers a range of fibers from widely used commercial fibers to developmental fibers. Consistent with previous radiation experiments, Nicalon fiber was severely degraded by the neutron irradiation. Similarly, Tyranno suffered severe degradation. The more advanced fibers which approach the composition and properties of SiC performed well under irradiation. Of these, the Carborundum SiC fiber appeared to perform the best. The Hi-Nicalon and Dow Corning Fibers exhibited good general stability, but also appear to have some surface roughening. The MER fibers and the Textron SCS-6 fibers both had carbon cores which adversely influenced the overall stability of the fibers.

  17. Retention of Hydrogen Isotopes in Neutron Irradiated Tungsten

    SciTech Connect

    Yuji Hatano; Masashi Shimada; Yasuhisa Oya; Guoping Cao; Makoto Kobayashi; Masanori Hara; Brad J. Merrill; Kenji Okuno; Mikhail A. Sokolov; Yutai Katoh

    2013-03-01

    To investigate the effects of neutron irradiation on hydrogen isotope retention in tungsten, disk-type specimens of pure tungsten were irradiated in the High Flux Isotope Reactor in Oak Ridge National Laboratory followed by exposure to high flux deuterium (D) plasma in Idaho National Laboratory. The results obtained for low dose n-irradiated specimens (0.025 dpa for tungsten) are reviewed in this paper. Irradiation at coolant temperature of the reactor (around 50 degrees C) resulted in the formation of strong trapping sites for D atoms. The concentrations of D in n-irradiated specimens were ranging from 0.1 to 0.4 mol% after exposure to D plasma at 200 and 500 degrees C and significantly higher than those in non-irradiated specimens because of D-trapping by radiation defects. Deep penetration of D up to a depth of 50-100 µm was observed at 500 degrees C. Release of D in subsequent thermal desorption measurements continued up to 900 degrees C. These results were compared with the behaviour of D in ion-irradiated tungsten, and distinctive features of n-irradiation were discussed.

  18. The physics experimental study for in-hospital neutron irradiator

    SciTech Connect

    Li Yiguo; Xia Pu; Zou Shuyun; Zhang Yongbao; Zheng Iv; Zheng Wuqing; Shi Yongqian; Gao Jijin; Zhou Yongmao

    2008-07-15

    MNSRs (Miniature Neutron Source Reactor) are low power research reactors designed and manufactured by China Institute of Atomic Energy (CIAE). MNSRs are mainly used for NAA, training and teaching, testing of nuclear instrumentation. The first MNSR, the prototype MNSR, was put into operation in 1984, later, eight other MNSRs had been built both at home and abroad. For MNSRs, highly enriched uranium (90%) is used as the fuel material. The In-Hospital Neutron Irradiator (IHNI) is designed for Boron Neutron Capture Therapy (BNCT) based on Miniature Neutron Source Reactor(MNSR). On both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the prompt gamma neutron activation analysis (PGNAA). In this paper, the experimental results of critical mass worth of the top Be reflectors worth of the control rod, neutron flux distribution and other components worth were measured, the experiment was done on the Zero Power Experiment equipment of MNSR. (author)

  19. Neutron-Irradiated Samples as Test Materials for MPEX

    DOE PAGESBeta

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of themore » samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.« less

  20. Neutron-Irradiated Samples as Test Materials for MPEX

    SciTech Connect

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.

  1. Development of positron annihilation spectroscopy for characterizing neutron irradiated tungsten

    NASA Astrophysics Data System (ADS)

    Taylor, C. N.; Shimada, M.; Merrill, B. J.; Drigert, M. W.; Akers, D. W.; Hatano, Y.

    2014-04-01

    Tungsten samples (6 mm diameter and 0.2 mm thick) were irradiated to 0.025 and 0.3 dpa with neutrons in the High Flux Isotope Reactor at Oak Ridge National Laboratory as part of the US/Japan Tritium, Irradiation and Thermofluids for America and Nippon (TITAN) collaboration. Samples were then exposed to deuterium plasma in Idaho National Laboratory's Tritium Plasma Experiment at 100, 200 and 500 °C to a total fluence of 1 × 1026 m-2. Nuclear reaction analysis (NRA) and Doppler broadening positron annihilation spectroscopy (DB-PAS) were performed at various stages to characterize radiation damage and retention. We present the first results of neutron irradiated tungsten characterized by DB-PAS in order to study defect concentration. Two positron sources, 22Na and 68Ge, probe ˜58 μm and through the entire 200 μm thick samples, respectively. DB-PAS results reveal clear differences between the various irradiated samples. These results, and a correlation between DB-PAS and NRA data, are presented.

  2. Radioactivity of neutron-irradiated cat's-eye chrysoberyls

    NASA Astrophysics Data System (ADS)

    Tang, S. M.; Tay, T. S.

    1999-04-01

    The recent report of marketing of radioactive chrysoberyl cat's-eyes in South-East Asian markets has led us to use an indirect method to estimate the threat to health these color-enhanced gemstones may pose if worn close to skin. We determined the impurity content of several cat's-eye chrysoberyls from Indian States of Orissa and Kerala using PIXE, and calculated the radioactivity that would be generated from these impurities and the constitutional elements if a chrysoberyl was irradiated by neutrons in a nuclear reactor for color enhancement. Of all the radioactive nuclides that could be created by neutron irradiation, only four ( 46Sc, 51Cr, 54Mn and 59Fe) would not have cooled down within a month after irradiation to the internationally accepted level of specific residual radioactivity of 2 nCi/g. The radioactivity of 46Sc, 51Cr and 59Fe would only fall to this safe limit after 15 months and that of 54Mn could remain above this limit for several years.

  3. 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.

  4. Patterned CoCrMo and Al2 O3 surfaces for reduced free wear debris in artificial joint arthroplasty.

    PubMed

    Tarabolsi, Mohamad; Klassen, Thomas; Mantwill, Frank; Gärtner, Frank; Siegel, Frank; Schulz, Arndt-Peter

    2013-12-01

    Surface wear of corresponding tribological pairings is still a major problem in the application of artificial joint surgery. This study aims at developing wear reduced surfaces to utilize them in total joint arthroplasty. Using a pico-second laser, samples of medical CoCrMo metal alloy and Al2 O3 ceramic were patterned by laser material removal. The subsequent tribological investigations employed a ring-on-disc method. The results showed that those samples with modified surfaces show less mass or volume loss than those with a regular, smooth surface. Using calf serum as lubricating medium, the volume loss of the structured CoCrMo samples was eight times lower than that of regular samples. By structuring Al2 O3 surfaces, the wear volume could be reduced by 4.5 times. The results demonstrate that defined surface channels or pits enable the local sedimentation of wear debris. Thus, the amount of free debris could be reduced. Fewer abrasives in the lubricated so-called three-body-wear between the contact surfaces should result in less surface damage. Apart from direct influences on the wear behavior, less amounts of free debris of artificial joints should also be beneficial for avoiding undesired reactions with the surrounding soft tissues. The results from this study are very promising. Future investigations should involve the use of simulators meeting the natural conditions in the joint and in vivo studies with living organisms. PMID:23595908

  5. Effect of proteins on the surface microstructure evolution of a CoCrMo alloy in bio-tribocorrosion processes.

    PubMed

    Wang, Zhongwei; Yan, Yu; Su, Yanjing; Qiao, Lijie

    2016-09-01

    Under tribological contact, the subsurface microstructure of CoCrMo alloys for artificial joint implants can be changed and affect the life and safety of such devices. As one of the most important and abundant components in the synovial fluid, proteins play a key role in affecting the bio-tribocorrosion behaviors of metal implants. The effect of proteins on the subsurface microstructure evolution of a CoCrMo alloy was investigated using a transmission electron microscope (TEM) in this study. The result shows that proteins have two main effects on the subsurface's evolution: forming a multilayered structure and causing severer subsurface deformation. The tribo-film can protect the passive film from scrapping, and then the passive film can reduce or even suppress the stacking fault annihilation by blocking the access to the metal surface. It leads to the stacking fault being diffused towards the deeper area and a strain accumulation in the subsurface, before inducing a severer deformation. On the other hand, the effect of proteins results in the location changing from the top surface to be underneath the top surface, where the maximum frictional shear stress occurs. This can cause a deeper deformation. PMID:27182652

  6. Evaluation of varying ductile fracture criteria for 42CrMo steel by compressions at different temperatures and strain rates.

    PubMed

    Quan, Guo-zheng; Luo, Gui-chang; Mao, An; Liang, Jian-ting; Wu, Dong-sen

    2014-01-01

    Fracturing by ductile damage occurs quite naturally in metal forming processes, and ductile fracture of strain-softening alloy, here 42CrMo steel, cannot be evaluated through simple procedures such as tension testing. Under these circumstances, it is very significant and economical to find a way to evaluate the ductile fracture criteria (DFC) and identify the relationships between damage evolution and deformation conditions. Under the guidance of the Cockcroft-Latham fracture criteria, an innovative approach involving hot compression tests, numerical simulations, and mathematic computations provides mutual support to evaluate ductile damage cumulating process and DFC diagram along with deformation conditions, which has not been expounded by Cockcroft and Latham. The results show that the maximum damage value appears in the region of upsetting drum, while the minimal value appears in the middle region. Furthermore, DFC of 42CrMo steel at temperature range of 1123~1348 K and strain rate of 0.01~10 s(-1) are not constant but change in a range of 0.160~0.226; thus, they have been defined as varying ductile fracture criteria (VDFC) and characterized by a function of temperature and strain rate. In bulk forming operations, VDFC help technicians to choose suitable process parameters and avoid the occurrence of fracture. PMID:24592175

  7. Methodologies for predicting the performance of Ni-Cr-Mo alloys proposed for high level nuclear waste containers

    SciTech Connect

    Dunn, D.S.; Cragnolino, G.A.; Sridhar, N.

    1999-07-01

    For the geologic disposal of the high level nuclear waste (HLW), aqueous corrosion is considered to be the most important factor in the long-term performance of containers, which are the main components of the engineered barrier subsystem. Container life, in turn, is important to the overall performance of the repository system. The proposed container designs and materials have evolved to include multiple barriers and highly corrosion resistant Ni-Cr-Mo alloys, such as Alloys 625 and C-22. Calculations of container life require knowledge of the initiation time and growth rate of localized corrosion. In the absence of localized corrosion, the rate of general or uniform dissolution, given by the passive current density of these materials, is needed. The onset of localized corrosion may be predicted by using the repassivation and corrosion potentials of the candidate container materials in the range of expected repository environments. In initial corrosion tests, chloride was identified as the most detrimental anionic species to the performance of Ni-Cr-Mo alloys. Repassivation potential measurements for Alloys 825, 625, and C-22, conducted over a wide range of chloride concentrations and temperatures, are reported. In addition, steady state passive current density, which will determine the container lifetime in the absence of localized corrosion, was measured for Alloy C-22 under various environmental conditions.

  8. Evolution of Morphology and Composition of the Carbides in Cr-Mo-V Steel after Service Exposure

    NASA Astrophysics Data System (ADS)

    Dong, Jiling; Shin, Keesam; He, Yinsheng; Song, Geewook; Jung, Jinesung

    2011-06-01

    Low alloy Cr-Mo-V steels are usually used in steam power generation units. The evolution of the carbides often leads to embrittlement of the components during elongated service. Therefore, the determination of carbide evolution mechanism during long-time service is important to understand and prevent premature failures such as temper embrittlement. In this study, low alloy Cr-Mo-V steels used as main steam pipes in a thermal power plant were studied after various service times as well as in the as-fabricated condition. Electron microscopic analyses were carried out on extraction replicas to observe and analyze the morphology and composition of the carbides. Predominant plate-like vanadium-rich carbides were observed in the as-fabricated condition. When exposed to on-site service, the V-rich carbides transformed to Mo-rich carbides which have a typical H morphology. The change of morphology and composition of the carbide is mainly due to the gradual depletion of Mo from the solid solution. In addition, a non-destructive carbide extraction method was established for examination of the precipitates in the working turbine rotor.

  9. Nanoarchitectured Co-Cr-Mo orthopedic implant alloys: nitrogen-enhanced nanostructural evolution and its effect on phase stability.

    PubMed

    Yamanaka, Kenta; Mori, Manami; Chiba, Akihiko

    2013-04-01

    Our previous studies indicate that nitrogen addition suppresses the athermal γ (face-centered cubic, fcc)→ε (hexagonal close-packed, hcp) martensitic transformation of biomedical Co-Cr-Mo alloys and ultimately offers large elongation to failure while maintaining high strength. In the present study, structural evolution and dislocation slip as an elementary process in the martensitic transformation in Co-Cr-Mo alloys were investigated to reveal the origin of their enhanced γ phase stability due to nitrogen addition. Alloy specimens with and without nitrogen addition were prepared. The N-doped alloys had a single-phase γ matrix, whereas the N-free alloys had a γ/ε duplex microstructure. Irrespective of the nitrogen content, dislocations frequently dissociated into Shockley partial dislocations with stacking faults. This indicates that nitrogen has little effect on the stability of the γ phase, which is also predicted by thermodynamic calculations. We discovered short-range ordering (SRO) or nanoscale Cr2N precipitates in the γ matrix of the N-containing alloy specimens, and it was revealed that both SRO and nanoprecipitates function as obstacles to the glide of partial dislocations and consequently significantly affect the kinetics of the γ→ε martensitic transformation. Since the formation of ε martensite plays a crucial role in plastic deformation and wear behavior, the developed nanostructural modification associated with nitrogen addition must be a promising strategy for highly durable orthopedic implants. PMID:23253619

  10. Evaluation of Varying Ductile Fracture Criteria for 42CrMo Steel by Compressions at Different Temperatures and Strain Rates

    PubMed Central

    Quan, Guo-zheng; Luo, Gui-chang; Mao, An; Liang, Jian-ting; Wu, Dong-sen

    2014-01-01

    Fracturing by ductile damage occurs quite naturally in metal forming processes, and ductile fracture of strain-softening alloy, here 42CrMo steel, cannot be evaluated through simple procedures such as tension testing. Under these circumstances, it is very significant and economical to find a way to evaluate the ductile fracture criteria (DFC) and identify the relationships between damage evolution and deformation conditions. Under the guidance of the Cockcroft-Latham fracture criteria, an innovative approach involving hot compression tests, numerical simulations, and mathematic computations provides mutual support to evaluate ductile damage cumulating process and DFC diagram along with deformation conditions, which has not been expounded by Cockcroft and Latham. The results show that the maximum damage value appears in the region of upsetting drum, while the minimal value appears in the middle region. Furthermore, DFC of 42CrMo steel at temperature range of 1123~1348 K and strain rate of 0.01~10 s−1 are not constant but change in a range of 0.160~0.226; thus, they have been defined as varying ductile fracture criteria (VDFC) and characterized by a function of temperature and strain rate. In bulk forming operations, VDFC help technicians to choose suitable process parameters and avoid the occurrence of fracture. PMID:24592175

  11. Manufacturing of high-strength Ni-free Co-Cr-Mo alloy rods via cold swaging.

    PubMed

    Yamanaka, Kenta; Mori, Manami; Yoshida, Kazuo; Kuramoto, Koji; Chiba, Akihiko

    2016-07-01

    The strengthening of biomedical metallic materials is crucial to increasing component durability in biomedical applications. In this study, we employ cold swaging as a strengthening method for Ni-free Co-Cr-Mo alloy rods and examine its effect on the resultant microstructures and mechanical properties. N is added to the alloy to improve the cold deformability, and a maximum reduction in area (r) of 42.6% is successfully obtained via cold swaging. The rod strength and ductility increase and decrease, respectively, with increasing cold-swaging reduction r. Further, the 0.2% proof stress at r=42.6% eventually reaches 1900MPa, which is superior to that obtained for the other strengthening methods proposed to date. Such significant strengthening resulting from the cold-swaging process may be derived from extremely large work hardening due to a strain-induced γ (fcc)→ε (hcp) martensitic transformation, with the resultant intersecting ε-martensite plates causing local strain accumulation at the interfaces. The lattice defects (dislocations/stacking faults) inside the ε phase also likely contribute to the overall strength. However, excessive application of strain during the cold-swaging process results in a severe loss in ductility. The feasibility of cold swaging for the manufacture of high-strength Co-Cr-Mo alloy rods is discussed. PMID:26773647

  12. Kinetics of borided 31CrMoV9 and 34CrAlNi7 steels

    SciTech Connect

    Efe, Goezde Celebi; Ipek, Mediha; Ozbek, Ibrahim; Bindal, Cuma

    2008-01-15

    In this study, kinetics of borides formed on the surface of 31CrMoV9 and 34CrAlNi7 steels borided in solid medium consisting of Ekabor II at 850-900-950 deg. C for 2, 4, 6 and 8 h were investigated. Scanning electron microscopy and optical microscopy examinations showed that borides formed on the surface of borided steels have columnar morphology. The borides formed in the coating layer confirmed by X-ray diffraction analysis are FeB, Fe{sub 2}B, CrB, and Cr{sub 2}B. The hardnesses of boride layers are much higher than that of matrix. It was found that depending on process temperature and time the fracture toughness of boride layers ranged from 3.93 to 4.48 MPa m{sup 1/2} for 31CrMoV9 and from 3.87 to 4.40 MPa m{sup 1/2} for 34CrAlNi7 steel. Activation energy, growth rate and growth acceleration of boride layer calculated according to these kinetic studies revealed that lower activation energy results in the fast growth rate and high growth acceleration.

  13. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    SciTech Connect

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.

  14. Long-term corrosion of Cr-Mo steels in superheated steam at 482 and 538/sup 0/C. [21/4 Cr-1 Mo; 9 Cr-1 Mo; Sumitomo 9 Cr-2 Mo; Sandvik HT-9

    SciTech Connect

    Griess, J.C.; DeVan, J.H.; Maxwell, W.A.

    1980-01-01

    The corrosion of several Cr-Mo ferritic steels was investigated in superheated steam at an operating power plant. Tests were conducted at 482 and 538/sup 0/C (900 and 1000/sup 0/F) in a once-through loop for times up to 28,000 h. Chromium concentrations ranged from 2.0 to 11.3%, and the effect of surface preparation on corrosion was investigated. Only one of many specimens showed evidence of exfoliation at 482/sup 0/C, but at 538/sup 0/C exfoliation occurred on at least some of the specimens of most materials; the exceptions were the alloy with the highest chromium content (Sandvik HT-9), one heat of 9 Cr-1 Mo steel with the highest silicon content, and Sumitomo 9 Cr-2 Mo steel, which was in test for only 19,000 h. Parabolic oxidation kinetics adequately described the corrosion process for about the first year, after which corrosion rates were constant and lower than predicted from extrapolation of the initial part of the penetration versus time curves. With chromium concentrations between 2 and 9%, corrosion behavior was independent of chromium content, and corrosion was only slightly less with Sandvik HT-9. Corrosion was nearly independent of surface preparation, but in two cases the presence of mill scale on the surface prior to steam exposure seemed to retard oxidation in steam. 11 figures, 5 tables.

  15. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Tobita, Tohru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-01

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 1024 n/m2 at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

  16. Effect of deep cryogenic treatment on the properties of 80CrMo12 5 tool steel

    NASA Astrophysics Data System (ADS)

    Amini, Kamran; Nategh, Said; Shafyei, Ali; Rezaeian, Ahmad

    2012-01-01

    The effect of deep cryogenic treatment on the mechanical properties of 80CrMo12 5 tool steel was investigated. Moreover, the effects of stabilization (holding at room temperature for some periods before deep cryogenic treatment) and tempering before deep cryogenic treatment were studied. The results show that deep cryogenic treatment can eliminate the retained austenite, making a better carbide distribution and a higher carbide amount. As a result, a remarkable improvement in wear resistance of cryogenically treated specimens is observed. Moreover, the ultimate tensile strength increases, and the toughness of the sample decreases. It is also found that both stabilization and tempering before deep cryogenic treatment decrease the wear resistance, hardness, and carbides homogeneity compared to the deep cryogenically treated samples. It is concluded that deep cryogenic treatment should be performed without any delay on samples after quenching to reach the highest wear resistance and hardness.

  17. Prediction of solidification path and carbide precipitation in Fe-C-V-Cr-Mo-W high speed steels

    NASA Astrophysics Data System (ADS)

    Zhang, Hongwei; Gandin, Charles-André; He, Jicheng; Nakajima, Keiji

    2012-07-01

    The solidification path and precipitation of carbides in the Fe-C-V-Cr-Mo-W high speed steel system are predicted with the help of thermodynamic equilibrium calculations. The Partial Equilibrium (PE) approximation is favoured. According to experimental data for high speed steel samples, the precipitating solidification sequence of carbides, including nature, composition and amount are discussed as a function of the nominal composition of C and V. The results show that the solidification path can be reasonably predicted by the Partial Equilibrium approximation for cooling rate lower than 10 K min-1. The experimental results suffer from the sensitivity limitation of the characterization methods used when the phase fraction becomes too small.

  18. Microstructural characterization of low and high carbon CoCrMo alloy nanoparticles produced by mechanical milling

    NASA Astrophysics Data System (ADS)

    Simoes, T. A.; Goode, A. E.; Porter, A. E.; Ryan, M. P.; Milne, S. J.; Brown, A. P.; Brydson, R. M. D.

    2014-06-01

    CoCrMo alloys are utilised as the main material in hip prostheses. The link between this type of hip prosthesis and chronic pain remains unclear. Studies suggest that wear debris generated in-vivo may be related to post-operative complications such as inflammation. These alloys can contain different amounts of carbon, which improves the mechanical properties of the alloy. However, the formation of carbides could become sites that initiate corrosion, releasing ions and/or particles into the human body. This study analysed the mechanical milling of alloys containing both high and low carbon levels in relevant biological media, as an alternative route to generate wear debris. The results show that low carbon alloys produce significantly more nanoparticles than high carbon alloys. During the milling process, strain induces an fcc to hcp phase transformation. Evidence for cobalt and molybdenum dissolution in the presence of serum was confirmed by ICP-MS and TEM EDX techniques.

  19. Room Temperature Microstructure and Property Evaluation of a Heat Treated Fully Bainitic 20CrMoVTiB410 Steel

    NASA Astrophysics Data System (ADS)

    Srivatsa, Kulkarni; Srinivas, Perla; Balachandran, G.; Balasubramanian, V.

    2016-08-01

    The room temperature mechanical behavior of the fully bainitic steel grade 20CrMoVTiB410 was studied in the as-quenched and tempered conditions. The hardenability response of the steel during heat treatment was assessed. In the as-quenched condition itself, the steel exhibited a good combination of strength, ductility and toughness. Tempering the quenched steel till to 550°C, showed uniform mechanical properties. Tempering at 650°C showed secondary hardening behaviour, where the highest strength and least impact toughness was observed. Tempering at 700°C showed a sharp decrease in strength but with significant enhancement of toughness. The properties obtained were correlated with the microstructure and phase analysis was established using optical, scanning electron microscope, transmission electron microscope and x-ray diffraction techniques.

  20. Study of the Effects of High Temperatures on the Engineering Properties of Steel 42CrMo4

    NASA Astrophysics Data System (ADS)

    Brnic, Josip; Turkalj, Goran; Canadija, Marko; Lanc, Domagoj; Brcic, Marino

    2015-02-01

    The paper presents and analyzes the experimental results of the effect of elevated temperatures on the engineering properties of steel 42CrMo4. Experimental data relating to the mechanical properties of the material, the creep resistance as well as Charpy impact energy. Temperature dependence of the mentioned properties is also shown. Some of creep curves were simulated using rheological models and an analytical equation. Finally, an assessment of fracture toughness was made that was based on experimentally determined Charpy impact energy. Based on the obtained results it is visible that the tensile strength (617 MPa) and yield strength (415 MPa) have the highest value at the room temperature while at the temperature of 700 °C (973 K) these values significantly decrease. This steel can be considered resistant to creep at 400 °C (673 K), but at higher temperatures this steel can be subjected to low levels of stress in a shorter time.

  1. Prediction of Mechanical Properties of 25CrMo48V Seamless Tube Using Neural Network Model

    NASA Astrophysics Data System (ADS)

    Sun, Laibo; Zhang, Chuanyou; Wang, Qingfeng; Wang, Mingzhi; Yan, Zesheng

    In this investigation, a neural network model was established to predict mechanical properties of 25CrMo48V seamless tubes. The sensitivity analysis was also performed to estimate the relative significance of each chemical composition in mechanical behavior of steel tubes. The results of this investigation show that there is a good agreement between experimental and predicted values indicating desirable validity of the model. Among those alloying elements, the elements of carbon, silicon and chromium tended to play a more important role in controlling both the yielding strength and the Charpy-V-Notch transverse impact toughness. In comparison, the impurities such as O, N, S and P have a relatively weak impact. More detailed dependences of mechanical properties on each chemical composition in isolation can be revealed using the established model. The well-trained neural network has a great potential in designing tough and ultrahigh-strength seamless tubes and modeling the on-line production parameters.

  2. Investigations on Freon-assisted atomization of refractory analytes (Cr, Mo, Ti, V) in multielement electrothermal atomic absorption spectrometry

    NASA Astrophysics Data System (ADS)

    Heinrich, Hans-Joachim; Matschat, Ralf

    2007-08-01

    Premixed 1% Freon in argon inner gas of various composition (CCl 2F 2, CHClF 2, CHF 3) was applied to graphite furnace atomizer to minimize unfavorable effects of carbide formation, such as signal tailing and memory effects in the simultaneous determination of Cr, Mo, Ti and V refractory analytes by electrothermal atomic absorption spectrometry using a multielement atomic absorption spectrometer. The effect of these gaseous additives was investigated when applied separately in atomization, pyrolysis and clean-out steps. The halogenation effects were analytically useful only under the precondition of using Ar-H 2 outer gas to the furnace to all heating steps, and also using this gas in the pre-atomization (drying, pyrolysis) steps. Optimum analytical performance was obtained when mixtures of 1% Freon in argon were applied just before and during the atomization step at a flow rate of 50 mL min - 1 and 2% hydrogen was used as purge gas. Using optimum conditions, signal tailings and carry-over contamination were reduced effectively and good precision (relative standard deviation below 1%) could be attained. Applying 1% CHClF 2 and an atomization temperature of 2550 °C, the characteristic masses obtained for simple aqueous solutions were 8.8 pg for Cr, 17 pg for Mo, 160 pg for Ti, and 74 pg for V. The limits of detection were 0.05, 0.2, 2.3 and 0.5 μg L - 1 for Cr, Mo, Ti and V, respectively. The developed method was applied to the analysis of digests of advanced ceramics. The accuracy of the procedure was confirmed by analyzing the certified reference material ERM-ED 102 (Boron Carbide Powder) and a silicon nitride powder distributed in the inter-laboratory comparison CCQM-P74.

  3. Study of cellular dynamics on polarized CoCrMo alloy using time-lapse live-cell imaging.

    PubMed

    Haeri, Morteza; Gilbert, Jeremy L

    2013-11-01

    The physico-chemical processes and phenomena occurring at the interface of metallic biomedical implants and the body dictate their successful integration in vivo. Changes in the surface potential and the associated redox reactions at metallic implants can significantly influence several aspects of biomaterial/cell interactions such as cell adhesion and survival in vitro. Accordingly, there is a voltage viability range (voltages which do not compromise cellular viability of the cells cultured on the polarized metal) for metallic implants. We report on cellular dynamics (size, polarity, movement) and temporal changes in the number and total area of focal adhesion complexes in transiently transfected MC3T3-E1 pre-osteoblasts cultured on CoCrMo alloy surfaces polarized at the cathodic and anodic edges of its voltage viability range (-400 and +500 mV (Ag/AgCl), respectively). Nucleus dynamics (size, circularity, movement) and the release of reactive oxygen species (ROS) were also studied on the polarized metal at -1000, -400 and +500 mV (Ag/AgCl). Our results show that at -400 mV, where reduction reactions dominate, a gradual loss of adhesion occurs over 24 h while cells shrink in size during this time. At +500 mV, where oxidation reactions dominate (i.e. metal ions form, including Cr6+), cells become non-viable after 5h without showing any significant changes in adhesion behavior right before cell death. Nucleus size of cells at -1000 mV decreased sharply within 15 min after polarization, which rendered the cells completely non-viable. No significant amount of ROS release by cells was detected on the polarized CoCrMo at any of these voltages. PMID:23831720

  4. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    SciTech Connect

    Hu, J. P.; Holden, N. E.; Reciniello, R. N.

    2014-05-23

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

  5. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    NASA Astrophysics Data System (ADS)

    Hu, J.-P.; Holden, N. E.; Reciniello, R. N.

    2016-02-01

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4-7% lower than

  6. 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.

  7. Progress on performance assessment of ITER enhanced heat flux first wall technology after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Hirai, T.; Bao, L.; Barabash, V.; Chappuis, Ph; Eaton, R.; Escourbiac, F.; Giqcuel, S.; Merola, M.; Mitteau, R.; Raffray, R.; Linke, J.; Loewenhoff, Th; Pintsuk, G.; Wirtz, M.; Boomstra, D.; Magielsen, A.; Chen, J.; Wang, P.; Gervash, A.; Safronov, V.

    2016-02-01

    ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6-0.8 dpa at 200 °C-250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m-2). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

  8. Behaviour of neutron irradiated beryllium during temperature excursions up to and beyond its melting temperature

    NASA Astrophysics Data System (ADS)

    Pajuste, Elina; Kizane, Gunta; Avotiņa, Līga; Zariņš, Artūrs

    2015-10-01

    Beryllium pebble behaviour has been studied regarding the accidental operation conditions of tritium breeding blanket of fusion reactors. Structure evolution, oxidation and thermal properties have been compared for nonirradiated and neutron irradiated beryllium pebbles during thermal treatment in a temperature range from ambient temperature to 1600 K. For neutron irradiated pebbles tritium release process was studied. Methods of temperature programmed tritium desorption (TPD) in combination with thermogravimetry (TG) and temperature differential analysis (TDA), scanning electron microscopy (SEM) in combination with Energy Dispersive X-ray analysis (EDX) have been used. It was found that there are strong relation between tritium desorption spectra and structural evolution of neutron irradiated beryllium. The oxidation rate is also accelerated by the structure damages caused by neutrons.

  9. Deuterium Depth Profile in Neutron-Irradiated Tungsten Exposed to Plasma

    SciTech Connect

    Masashi Shimada; G. Cao; Y. Hatano; T. Oda; Y. Oya; M. Hara; P. Calderoni

    2011-05-01

    The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

  10. The Influence of Composition upon Surface Degradation and Stress Corrosion Cracking of the Ni-Cr-Mo Alloys in Wet Hydrofluoric Acid

    SciTech Connect

    Crook, P; Meck, N S; Rebak, R B

    2006-12-04

    At concentrations below 60%, wet hydrofluoric acid (HF) is extremely corrosive to steels, stainless steels and reactive metals, such as titanium, zirconium, and tantalum. In fact, only a few metallic materials will withstand wet HF at temperatures above ambient. Among these are the nickel-copper (Ni-Cu) and nickel-chromium-molybdenum (Ni-Cr-Mo) alloys. Previous work has shown that, even with these materials, there are complicating factors. For example, under certain conditions, internal attack and stress corrosion cracking (SCC) are possible with the Ni-Cr-Mo alloys, and the Ni-Cu materials can suffer intergranular attack when exposed to wet HF vapors. The purpose of this work was to study further the response of the Ni-Cr-Mo alloys to HF, in particular their external corrosion rates, susceptibility to internal attack and susceptibility to HF-induced SCC, as a function of alloy composition. As a side experiment, one of the alloys was tested in two microstructural conditions, i.e. solution annealed (the usual condition for materials of this type) and long-range ordered (this being a means of strengthening the alloy in question). The study of external corrosion rates over wide ranges of concentration and temperature revealed a strong beneficial influence of molybdenum content. However, tungsten, which is used as a partial replacement for molybdenum in some Ni-Cr-Mo alloys, appears to render the alloys more prone to internal attack. With regard to HF-induced SCC of the Ni-Cr-Mo alloys, this study suggests that only certain alloys (i.e., those containing tungsten) exhibit classical SCC. It was also discovered that high external corrosion rates inhibit HF-induced SCC, presumably due to rapid progression of the external attack front. With regard to the effects of long-range ordering, these were only evident at the highest test temperatures, where the ordered structure exhibited much higher external corrosion rates than the annealed structure.

  11. Migration and accumulation at dislocations of transmutation helium in austenitic steels upon neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kozlov, A. V.; Portnykh, I. A.

    2016-04-01

    The model of the migration and accumulation at dislocations of transmutation helium and the formation of helium-vacancy pore nuclei in austenitic steels upon neutron irradiation has been proposed. As illustrations of its application, the dependences of the characteristics of pore nuclei on the temperature of neutron irradiation have been calculated. The results of the calculations have been compared with the experimental data in the literature on measuring the characteristics of radiation-induced porosity that arises upon the irradiation of shells of fuel elements of a 16Cr-19Ni-2Mo-2Mn-Si-Ti-Nb-V-B steel in a fast BN600 neutron reactor at different temperatures.

  12. Effects of hole doping by neutron irradiation of magnetic field induced electronic phase transitions in graphite

    SciTech Connect

    Singleton, John; Yaguchi, Hiroshi

    2008-01-01

    We have investigated effects of hole doping by fast-neutron irradiation on the magnetic-field induced phase transitions in graphite using specimens irradiated with fast neutrons. Resistance measurements have been done in magnetic fields of up to above 50 T and at temperatures down to about 1.5 K. The neutron irradiation creates lattice defects acting as acceptors, affecting the imbalance of the electron and hole densities and the Fermi level. We have found that the reentrant field from the field induced state back to the normal state shifts towards a lower field with hole doping, suggestive of the participation of electron subbands in the magnetic-field induced state.

  13. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    PubMed

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard. PMID:21071234

  14. Location and chemical bond of radionuclides in neutron-irradiated nuclear graphite

    NASA Astrophysics Data System (ADS)

    Vulpius, D.; Baginski, K.; Fischer, C.; Thomauske, B.

    2013-07-01

    The locations and the chemical forms (chemical bonds) of radionuclides in neutron-irradiated nuclear graphite have been determined in order to develop principal strategies for the management of graphitic nuclear waste. Due to the relatively low concentration of radionuclides in neutron-irradiated nuclear graphite (<1 ppm) direct spectroscopic methods are not applicable to investigate chemical structures. Therefore, methods by analogy have been applied. Such methods are investigations of the chemically detectable precursors of radionuclides in neutron-irradiated nuclear graphite and subjection of irradiated graphite to different chemical reactions followed by measurements of the radionuclide-containing reaction products by sensitive radiochemical methods. The paper discusses the applicability of these methods. The radionuclides investigated in this study can be divided into three parts: tritium, radiocarbon and metallic activation and fission products. Tritium can be bound in neutron-irradiated nuclear graphite as strongly adsorbed tritiated water (HTO), in oxygen-containing functional groups (e.g. C-OT) and as hydrocarbons (C-T). Radiocarbon is covalently bound with the graphite structure. The activity can be described by a homogeneously distributed part and a heterogeneously distributed part (enriched on surfaces or in hotspots). Metallic radionuclides can be bound as ions or covalent metal-carbon compounds. The distribution of all these radionuclides is mainly dependent on the distribution of their inactive precursors.

  15. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    SciTech Connect

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

  16. Comparison of Deuterium Retention for Ion-irradiated and Neutron-irradiated Tungsten

    SciTech Connect

    Yasuhisa Oya; Masashi Shimada; Makoto Kobayashi; Takuji Oda; Masanori Hara; Hideo Watanabe; Yuji Hatano; Pattrick Calderoni; Kenji Okuno

    2011-12-01

    The behavior of D retention for Fe{sup 2+}-irradiated tungsten with a damage of 0.025-3 dpa was compared with that for neutron-irradiated tungsten with 0.025 dpa. The D{sub 2} thermal desorption spectroscopy (TDS) spectra for Fe{sup 2+}-irradiated tungsten consisted of two desorption stages at 450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an addition desorption stage was found at 750 K. The desorption rate of the major desorption stage at 550K increased as the displacement damage increased due to Fe{sup 2+} irradiation increasing. In addition, the first desorption stage at 450K was found only for damaged samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced by Fe{sup 2+} irradiation, and the first stage should be the accumulation of D in mono-vacancy and the activation energy would be relatively reduced, where the dislocation loop and vacancy is produced. The third one was found only for neutron irradiation, showing the D trapping by a void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at half-maximum of the TDS spectrum. Therefore, it can be said that the D{sub 2} TDS spectra for Fe{sup 2+}-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different from that for the ion-irradiated one.

  17. Mechanical and thermal properties of hot pressed CoCrMo-porcelain composites developed for prosthetic dentistry.

    PubMed

    Henriques, B; Gasik, M; Souza, J C M; Nascimento, R M; Soares, D; Silva, F S

    2014-02-01

    In this study, mechanical and thermal properties of CoCrMo-porcelain composites for dental restorations have been evaluated. These metal-ceramic composites were produced by powder metallurgy and hot pressing techniques from the mixtures of metal and ceramic powders with different volume fractions. Young's moduli and the coefficient of thermal expansion of materials were evaluated by dynamic mechanical analysis (DMA) and dilatometry (DIL) tests, respectively. The strength in flexion and shear was measured with a universal test machine and hardness with a respective tester. The microstructures and fracture surfaces were inspected by the means of optical microscopy and Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS). Shear strength, Flexural strength and Young' moduli of ceramic and metal-matrix composites were found to increase with higher metal particles content. The DMA tests performed at different frequencies showed no frequency-dependent features of the materials studied, indicating no viscoelastic behavior. The fracture surfaces analysis suggests the load-transfer mechanism be possibly responsible for this behavior, as the differences in CTE are low enough to cause significant thermal stresses in these materials. The results might be included in a materials properties database for further use for design and optimization of dental restorations. PMID:24269945

  18. Elevated temperature creep-fatigue crack propagation in nickel-base alloys and 1 Cr-Mo-V steel

    NASA Astrophysics Data System (ADS)

    Nazmy, M.; Hoffelner, W.; Wüthrich, C.

    1988-04-01

    The crack growth behavior of several high temperature nickel-base alloys, under cyclic and static loading, is studied and reviewed. In the oxide dispersion strengthened (ODS) MA 6000 and MA 754 alloys, the high temperature crack propagation exhibited orientation dependence under cyclic as well as under static loading. The creep crack growth (CCG) behavior of cast nickel-base IN-738 and IN-939* superalloys at 850 °C could be characterized by the stress intensity factor, K 1. In the case of the alloy IN-901 at 500 °C and 600 °C, K 1 was found to be the relevant parameter to characterize the creep crack growth behavior. The energy rate line integral, C*, may be the appropriate loading parameter to describe the creep crack growth behavior of the nickel-iron base IN-800H alloy at 800 °C. The creep crack growth data of 1 Cr-Mo-V steel, with bainitic microstructure, at 550 °C could be correlated better by C * than by K 1.

  19. Tearing Resistance Properties of Cr-Mo Steels with Internal Hydrogen Determined by the Potential Drop Method

    NASA Astrophysics Data System (ADS)

    Konosu, Shinji; Shimazu, Hidenori; Fukuda, Ryohei

    2015-12-01

    The tearing resistance, dJ/da, of conventional 2.25Cr-1Mo steels and a V-bearing steel (2.25Cr-1Mo-0.3V steel) with internal hydrogen was measured using the effective offset potential drop method. Internal hydrogen refers to test specimens that are precharged (thermally charged) prior to testing. In general, Cr-Mo steels, used widely in the refining and petrochemical industries, are susceptible to temper embrittlement. However, very few studies have dealt with the effects of hydrogen and temper embrittlement on the tearing resistance. Test specimens were prepared by subjecting them to normalizing, tempering, and post-weld heat treatments that simulated actual conditions. Some specimens were embrittled by step cooling. Hydrogen substantially reduced dJ/da for all samples except for that for the V-bearing steel, and temper embrittlement caused additional adverse effects on dJ/da for samples with internal hydrogen for which the temper embrittlement parameter, i.e., the J-factor, was large.

  20. Nd:YAG laser cladding of Co-Cr-Mo alloy on γ-TiAl substrate

    NASA Astrophysics Data System (ADS)

    Barekat, Masoud; Shoja Razavi, Reza; Ghasemi, Ali

    2016-06-01

    In this work, Co-Cr-Mo powder is used to form laser clads on a γ-TiAl substrate. The single-track geometrical characteristics such as width, height, penetration depth, dilution and wetting angle play the important role to control the characteristics of laser clad coatings formed by overlap of individual tracks. This paper is investigated the relations between the main coaxial laser cladding parameters (laser power P, laser beam scanning speed S and powder feeding rate F) and geometrical characteristics of single tracks by linear regression analysis. The results show that the clad height, H, depends linearly on the FS-5/4 parameter with the laser power having a minimal effect. Similarly, the cladding width W is controlled by the PS-2/3 parameter. The penetration depth b and dilution, D are proportional to P2S-1/4F-1/4 and P2/3S1/2F-1/2 respectively and wetting angle is controlled by the P1/4S1/2F-1/2 parameter. These empirical dependencies are observed with high values of the correlation coefficient (R>0.9). Finally, based on these relations, a laser cladd processing map was designed to use as a guideline for the selection of proper processing parameters for a required coating.

  1. Autophagy mediated CoCrMo particle-induced peri-implant osteolysis by promoting osteoblast apoptosis

    PubMed Central

    Wang, Zhenheng; Liu, Naicheng; Liu, Kang; Zhou, Gang; Gan, Jingjing; Wang, Zhenzhen; Shi, Tongguo; He, Wei; Wang, Lintao; Guo, Ting; Bao, Nirong; Wang, Rui; Huang, Zhen; Chen, Jiangning; Dong, Lei; Zhao, Jianning; Zhang, Junfeng

    2015-01-01

    Wear particle-induced osteolysis is the leading cause of aseptic loosening, which is the most common reason for THA (total hip arthroplasty) failure and revision surgery. Although existing studies suggest that osteoblast apoptosis induced by wear debris is involved in aseptic loosening, the underlying mechanism linking wear particles to osteoblast apoptosis remains almost totally unknown. In the present study, we investigated the effect of autophagy on osteoblast apoptosis induced by CoCrMo metal particles (CoPs) in vitro and in a calvarial resorption animal model. Our study demonstrated that CoPs stimulated autophagy in osteoblasts and PIO (particle-induced osteolysis) animal models. Both autophagy inhibitor 3-MA (3-methyladenine) and siRNA of Atg5 could dramatically reduce CoPs-induced apoptosis in osteoblasts. Further, inhibition of autophagy with 3-MA ameliorated the severity of osteolysis in PIO animal models. Moreover, 3-MA also prevented osteoblast apoptosis in an antiautophagic way when tested in PIO model. Collectively, these results suggest that autophagy plays a key role in CoPs-induced osteolysis and that targeting autophagy-related pathways may represent a potential therapeutic approach for treating particle-induced peri-implant osteolysis. PMID:26566231

  2. Microstructural evolution and mechanical properties of biomedical Co-Cr-Mo alloy subjected to high-pressure torsion.

    PubMed

    Isik, Murat; Niinomi, Mitsuo; Cho, Ken; Nakai, Masaaki; Liu, Huihong; Yilmazer, Hakan; Horita, Zenji; Sato, Shigeo; Narushima, Takayuki

    2016-06-01

    The effects of severe plastic deformation through high-pressure torsion (HPT) on the microstructure and tensile properties of a biomedical Co-Cr-Mo (CCM) alloy were investigated. The microstructure was examined as a function of torsional rotation number, N and equivalent strain, εeq in the HPT processing. Electron backscatter diffraction analysis (EBSD) shows that a strain-induced martensitic transformation occurs by the HPT processing. Grain diameter decreases with increasing εeq, and the HPT-processed alloy (CCMHPT) for εeq=45 exhibits an average grain diameter of 47nm, compared to 70μm for the CCM alloy before HPT processing. Blurred and wavy grain boundaries with low-angle of misorientation in the CCMHPT sample for εeq<45 become better-defined grain boundaries with high-angle of misorientation after HPT processing for εeq=45. Kernel average misorientation (KAM) maps from EBSD indicate that KAM inside grains increases with εeq for εeq<45, and then decreases for εeq=45. The volume fraction of the ε (hcp) phase in the CCMHPT samples slightly increases at εeq=9, and decreases at εeq=45. In addition, the strength of the CCMHPT samples increases at εeq=9, and then decrease at εeq=45. The decrease in the strength is attributed to the decrease in the volume fraction of ε phase, annihilation of dislocations, and decrease in strain in the CCMHPT sample processed at εeq=45 by HPT. PMID:26774617

  3. Does surface wettability influence the friction and wear of large-diameter CoCrMo alloy hip resurfacings?

    PubMed

    Curran, Sarah; Hoskin, Tom; Williams, Sarah; Scholes, Susan C; Kinbrum, Amy; Unsworth, Anthony

    2013-08-01

    The role of surface tension in the lubrication of metal-on-metal (CoCrMo alloy) hip resurfacings has been investigated to try to explain why all metal joints fail to be lubricated with simple water-based lubricants (sodium carboxymethyl cellulose), which have similar rheology to synovial fluid, but are lubricated with the same fluid with the addition of a proportion of bovine serum. As part of this study, surfactants, in the form of detergents, when added to carboxymethyl cellulose, have been shown to produce a predominantly fluid-film lubrication mechanism with friction even lower than the biological lubricant containing serum. Friction factors were reduced by 80% when a detergent was added to the lubricant. It is considered that the failure of the water-based fluids to generate fluid-film lubrication is due to the fact that 'boundary slip' takes place where the fluid does not fully attach to the bounding solid surfaces as assumed in Reynolds' equation, thereby drawing in less lubricant than predicted from hydrodynamic theory. The addition of surfactants either in the form of natural materials such as serum or in the form of detergent reduces surface tension and helps the water-based lubricant to attach more fully to the bounding surfaces resulting in more fluid entrainment and thicker fluid-film formation. This was confirmed by up to 70% lower wear being found when these joints were lubricated in a detergent solution rather than 25% bovine serum. PMID:23852389

  4. High temperature passive film on the surface of Co-Cr-Mo alloy and its tribological properties

    NASA Astrophysics Data System (ADS)

    Guo, Feifei; Dong, Guangneng; Dong, Lishe

    2014-09-01

    For the artificial hip joints, passive film formed on the Co-Cr-Mo alloy acted as a highly protective barrier in the body fluid. But its stability, composition and structure always influenced the protection. In this work, passive film was obtained by high temperature treatment. The effect of passivation environment on the properties of the passive film was investigated. The film's surface roughness, micro-hardness and structure were analyzed. In order to study the tribological behavior of the passive film, pin-on-disk tribotest was carried out under bovine serum albumin (BSA) and saline solution. Results indicated the sample passivated in vacuum had friction coefficient of 0.18 under BSA solution and 0.53 under saline solution; the sample passivated in air had friction coefficient of 0.14 under BSA solution and 0.56 under saline solution. In addition, the reference sample without passivation was tested under the same condition. It showed friction of 0.22 under BSA solution and 0.45 under solution. The lubricating mechanism was attributed to BSA tribo-film absorption on the surface and high hardness passive film.

  5. Creep deformation and fracture of a Cr/Mo/V bolting steel containing selected trace-element additions

    NASA Astrophysics Data System (ADS)

    Larouk, Z.; Pilkington, R.

    1999-08-01

    The article reports the creep behavior, at 565 °C, of 1Cr1Mo0.75V (Ti, B) (Durehete D1055) steel, in each of two grain sizes and doped with individual trace elements such as P, As, and Sn, in comparison to a reference cast of the base material containing 0.08 wt pct Ti. The addition of the trace elements P, As, or Sn (each <0.045 wt pct) appears to produce no significant effect on creep strength or creep crack-growth resistance at 565 °C. The fine-grained material shows low creep strength but notch strengthening, while the coarse-grained material shows higher creep strength and exhibits notch weakening for test times up to 2750 hours. From creep crack-growth tests, it appears that the C* parameter is not appropriate for correlating the creep crack-growth rate under the present test conditions. The parameters K I or σ net are found to correlate better, but, from the present data, it is not possible to judge which of these parameters is more appropriate for general use. It is suggested that the presence of Ti in CrMoV steels has an inhibiting effect on trace-element embrittlement.

  6. Characterization of precipitates in X12CrMoWVNbN10-1-1 steel during heat treatment

    NASA Astrophysics Data System (ADS)

    Tao, Xingang; Gu, Jianfeng; Han, Lizhan

    2014-09-01

    The characterization of precipitates in X12CrMoWVNbN10-1-1 steel during the heat treatment was carried out for revealing the evolution of the precipitates. In addition to other microstructural parameters (such as dislocation and subgrains), the precipitate also plays an important role for microstructural stability which is a prerequisite for long term creep strength. In this paper, the precipitates during the heat treatment for this steel were characterized using physicochemical phase analyses and transmission electron microscopy. It was found that the Fe-rich M3C carbides and Nb-rich MX particles were detected in the samples cooled in furnace from austenitization at 1080 °C for 16 h. However, after water cooling, only Nb-rich MX particles existed. During tempering at 570 °C for 18 h, the formation of Cr-rich M7C3 was detected but was replaced partially by Cr-rich M23C6. Additional Cr-rich M2N nitride was also found. After two successive tempering (570 °C + 690 °C) for 24 h, Cr-rich M7C3 was completely replaced. The microchemical analyses of the extracted residues during heat treatment were also discussed. The results gave rise to an indication that the precipitation of precipitates nearly completed in first tempering and the transformation from Cr-rich M7C3 to Cr-rich M23C6 mainly occurred in the second tempering.

  7. Increased osteoblast adhesion on nanophase metals: Ti, Ti6Al4V, and CoCrMo.

    PubMed

    Webster, Thomas J; Ejiofor, Jeremiah U

    2004-08-01

    Previous studies have demonstrated increased functions of osteoblasts (bone-forming cells) on nanophase compared to conventional ceramics (specifically, alumina, titania, and hydroxyapatite), polymers (such as poly lactic-glycolic acid and polyurethane), carbon nanofibers/nanotubes, and composites thereof. Nanophase materials are unique materials that simulate dimensions of constituent components of bone since they possess particle or grain sizes less than 100 nm. However, to date, interactions of osteoblasts on nanophase compared to conventional metals remain to be elucidated. For this reason, the objective of the present in vitro study was to synthesize, characterize, and evaluate osteoblast adhesion on nanophase metals (specifically, Ti, Ti6Al4V, and CoCrMo alloys). Such metals in conventional form are widely used in orthopedic applications. Results of this study provided the first evidence of increased osteoblast adhesion on nanophase compared to conventional metals. Interestingly, osteoblast adhesion occurred preferentially at surface particle boundaries for both nanophase and conventional metals. Since more particle boundaries are present on the surface of nanophase compared to conventional metals, this may be an explanation for the measured increased osteoblast adhesion. Lastly, material characterization studies revealed that nanometal surfaces possessed similar chemistry and only altered in degree of nanometer surface roughness when compared to their respective conventional counterparts. Because osteoblast adhesion is a necessary prerequisite for subsequent functions (such as deposition of calcium-containing mineral), the present study suggests that nanophase metals should be further considered for orthopedic implant applications. PMID:15120519

  8. Temperature-dependent phase-specific deformation mechanisms in a directionally solidified NiAl-Cr(Mo) lamellar composite

    DOE PAGESBeta

    Yu, Dunji; An, Ke; Chen, Xu; Bei, Hongbin

    2015-10-09

    Phase-specific thermal expansion and mechanical deformation behaviors of a directionally solidified NiAl–Cr(Mo) lamellar in situ composite were investigated by using real-time in situ neutron diffraction during compression at elevated temperatures up to 800 °C. Tensile and compressive thermal residual stresses were found to exist in the NiAl phase and Crss (solid solution) phase, respectively. Then, based on the evolution of lattice spacings and phase stresses, the phase-specific deformation behavior was analyzed qualitatively and quantitatively. Moreover, estimates of phase stresses were derived by Hooke's law on the basis of a simple method for the determination of stress-free lattice spacing in inmore » situ composites. During compressive loading, the NiAl phase yields earlier than the Crss phase. The Crss phase carries much higher stress than the NiAl phase, and displays consistent strain hardening at all temperatures. The NiAl phase exhibits strain hardening at relatively low temperatures and softening at high temperatures. During unloading, the NiAl phase yields in tension whereas the Crss phase unloads elastically. Additionally, post-test microstructural observations show phase-through cracks at room temperature, micro cracks along phase interfaces at 600 °C and intact lamellae kinks at 800 °C, which is due to the increasing deformability of both phases as temperature rises.« less

  9. Examination of Corrosion Products and the Alloy Surface After Crevice Corrosion of a Ni-Cr-Mo- Alloy

    SciTech Connect

    X. Shan; J.H. Payer

    2006-06-09

    The objective of this study is to investigate the composition of corrosion products and the metal surface within a crevice after localized corrosion. The analysis provides insight into the propagation, stifling and arrest processes for crevice corrosion and is part of a program to analyze the evolution of localized corrosion damage over long periods of time, i.e. 10,000 years and longer. The approach is to force the initiation of crevice corrosion by applying anodic polarization to a multiple crevice assembly (MCA). Results are reported here for alloy C-22, a Ni-Cr-Mo alloy, exposed to a high temperature, concentrated chloride solution. Controlled crevice corrosion tests were performed on C-22 under highly aggressive, accelerated condition, i.e. 4M NaCl, 100 C and anodic polarization to -0.15V-SCE. The crevice contacts were by either a polymer tape (PTFE) compressed by a ceramic former or by a polymer (PTFE) crevice former. Figure 1 shows the polarization current during a crevice corrosion test. After an incubation period, several initiation-stifle-arrest events were indicated. The low current at the end of the test indicated that the metal surface had repassivated.

  10. Reduced Pressure Electron Beam Welding Evaluation Activities on a Ni-Cr-Mo Alloy for Nuclear Waste Packages

    SciTech Connect

    Wong, F; Punshon, C; Dorsch, T; Fielding, P; Richard, D; Yang, N; Hill, M; DeWald, A; Rebak, R; Day, S; Wong, L; Torres, S; McGregor, M; Hackel, L; Chen, H-L; Rankin, J

    2003-09-11

    The current waste package design for the proposed repository at Yucca Mountain Nevada, USA, employs gas tungsten arc welding (GTAW) in fabricating the waste packages. While GTAW is widely used in industry for many applications, it requires multiple weld passes. By comparison, single-pass welding methods inherently use lower heat input than multi-pass welding methods which results in lower levels of weld distortion and also narrower regions of residual stresses at the weld TWI Ltd. has developed a Reduced Pressure Electron Beam (RPEB) welding process which allows EB welding in a reduced pressure environment ({le} 1 mbar). As it is a single-pass welding technique, use of RPEB welding could (1) achieve a comparable or better materials performance and (2) lead to potential cost savings in the waste package manufacturing as compared to GTAW. Results will be presented on the initial evaluation of the RPEB welding on a Ni-Cr-Mo alloy (a candidate alloy for the Yucca Mountain waste packages) in the areas of (a) design and manufacturing simplifications, (b) material performance and (c) weld reliability.

  11. Comparison of Crevice Corrosion of Fe-Based Amorphous Metal and Crystalline Ni-Cr-Mo Alloy

    SciTech Connect

    Shan, X; Ha, H; Payer, J H

    2008-07-24

    The crevice corrosion behaviors of an Fe-based bulk metallic glass alloy (SAM1651) and a Ni-Cr-Mo crystalline alloy (C-22) were studied in 4M NaCl at 100 C with cyclic potentiodynamic polarization and constant potential tests. The corrosion damage morphologies, corrosion products and the compositions of corroded surfaces of these two alloys were studied with optical 3D reconstruction, Scanning Electron Microscopy (SEM), Energy Dispersive Spectroscopy (EDS) and Auger Electron Spectroscopy (AES). It was found that the Fe-based bulk metallic glass (amorphous alloy) SAM1651 had a more positive breakdown potential and repassivation potential than crystalline alloy C-22 in cyclic potentiodynamic polarization tests and required a more positive oxidizing potential to initiate crevice corrosion in constant potential test. Once crevice corrosion initiated, the corrosion propagation of C-22 was more localized near the crevice border compared to SAM1651, and SAM1651 repassivated more readily than C-22. The EDS results indicated that the corrosion products of both alloys contained high amount of O and were enriched in Mo and Cr. The AES results indicated that a Cr-rich oxide passive film was formed on the surfaces of both alloys, and both alloys were corroded congruently.

  12. Effect of tribolayer formation on corrosion of CoCrMo alloys investigated using scanning electrochemical microscopy.

    PubMed

    Meyer, Joshua N; Mathew, Mathew T; Wimmer, Markus A; LeSuer, Robert J

    2013-08-01

    Scanning electrochemical microscopy was used to probe the topography and electrochemical activity of CoCrMo alloys mechanically polished in the presence of bovine calf serum (BCS) in a hip simulator. These substrates are made of the same alloy used in metal-on-metal bearings for artificial hip joints. The BCS serves as an in vitro substitute for the synovial fluid which forms a lubricant in the actual orthopedic device. Chemical and mechanical processes result in the formation of a tribolayer which passivates the alloy surface. Our studies of the heterogeneous electron transfer between ferrocenemethanol and the alloy indicate that the tribolayer formed on both high- and low-carbon substrates is highly heterogeneous with regions of high electrochemical activity. Whereas pits in the samples polished in the absence of BCS show the regions of highest electrochemical activity, the tribolayer-coated samples have electrochemical hot spots in topographically smooth regions of the surface. The tribolayer provides some attenuation of the electrochemical activity of the alloy but does not prevent the possibility of corrosion from occurring. PMID:23848566

  13. The effect of porous coating processing on the corrosion behavior of cast Co-Cr-Mo surgical implant alloys.

    PubMed

    Jacobs, J J; Latanision, R M; Rose, R M; Veeck, S J

    1990-11-01

    The manufacture of porous coated cobalt-based surgical implant alloys requires sintering--a high temperature process above the incipient melting temperature of this alloy system. The metallurgical changes produced by the high temperature sinter cycle consist of dissolution of interdendritic carbides, massive precipitation of lamellar carbide eutectic phases at grain boundaries, localized porosity from incipient melting that is not completely eliminated by subsequent hot isostatic pressing, and grain growth in fine-grained materials. These microstructural changes, which are known to affect the mechanical properties, do not affect the static in vitro localized and generalized corrosion behavior of the bulk material as determined by anodic polarization measurements in a buffered saline environment and direct examination by scanning electron and optical microscopy. Additionally, cast Co-Cr-Mo surgical implant alloys are found to be immune to crevice corrosion (in the absence of mechanical fretting) in the saline environment studied. The hysteretic component of the anodic polarization curve is not due to crevice corrosion; rather, as suggested by the electrochemical tests and Auger spectroscopy, the hysteresis is due to redox reactions in the chromium-rich surface layer. PMID:2213344

  14. Temperature-dependent phase-specific deformation mechanisms in a directionally solidified NiAl-Cr(Mo) lamellar composite

    SciTech Connect

    Yu, Dunji; An, Ke; Chen, Xu; Bei, Hongbin

    2015-10-09

    Phase-specific thermal expansion and mechanical deformation behaviors of a directionally solidified NiAl–Cr(Mo) lamellar in situ composite were investigated by using real-time in situ neutron diffraction during compression at elevated temperatures up to 800 °C. Tensile and compressive thermal residual stresses were found to exist in the NiAl phase and Crss (solid solution) phase, respectively. Then, based on the evolution of lattice spacings and phase stresses, the phase-specific deformation behavior was analyzed qualitatively and quantitatively. Moreover, estimates of phase stresses were derived by Hooke's law on the basis of a simple method for the determination of stress-free lattice spacing in in situ composites. During compressive loading, the NiAl phase yields earlier than the Crss phase. The Crss phase carries much higher stress than the NiAl phase, and displays consistent strain hardening at all temperatures. The NiAl phase exhibits strain hardening at relatively low temperatures and softening at high temperatures. During unloading, the NiAl phase yields in tension whereas the Crss phase unloads elastically. Additionally, post-test microstructural observations show phase-through cracks at room temperature, micro cracks along phase interfaces at 600 °C and intact lamellae kinks at 800 °C, which is due to the increasing deformability of both phases as temperature rises.

  15. Restoration of Obliterated Numbers on 40NiCrMo4 Steel by Etching Method: Metallurgical and Statistical Approaches.

    PubMed

    Fortini, Annalisa; Merlin, Mattia; Soffritti, Chiara; Garagnani, Gian L

    2016-01-01

    The restoration of obliterated serial numbers is a problem of common occurrence in the forensic field. Among several restoration techniques, chemical etching is the most frequently used. The present research is aimed at studying the restoration of serial numbers, stamped on 40NiCrMo4 steel plates, by means of chemical etching. Microstructural characterization was firstly carried out to study the plastically deformed regions surrounding the marks. The obliteration was performed by controlled removals of material at increasing depths of erasure, and five etching reagents were considered to analyze their sensitivity and effectiveness. Experimental results revealed that Fry's reagent was the most sensitive, able to restore erased marks up to 60 μm under the depth of the imprint. The reagent comprising 25 mL HNO3 and 75 mL H2 O provided good results, recovering the major numbers of characters. A descriptive statistical analysis was conducted to study the operator's influence on the recovered marks' identification. PMID:26250339

  16. The new 7CrMoVTiB10-10 (T 24) material for boiler waterwalls

    SciTech Connect

    Husemann, R.U.; Bendick, W.; Haarmann, K.

    1999-07-01

    Besides the main components such as headers, piping, superheaters and separators, the waterwalls as flue gas-tight enclosure of the combustion chamber is of central importance for the design of a steam generator. The safety margins are more and more exhausted for the conventional materials, used so far for waterwalls, by increasing operating pressures and temperatures to raise the efficiency of a power plant. The former safety margin of the waterwall materials might be recovered by the new material T 24 or 7CrMoVTiB10-10 (2.5Cr, 1Mo, V, Ti, B). For this material creep rupture test results for test periods of up to about 100.000 h are already available. Test results of the base metal for tubes and pipes and the German qualification program which has been running for 4 years will be presented which includes qualification tests for TIG circumferential welds, SAW welds, creep rupture tests on welds, aging test on weld metal and test applications in power stations.

  17. EFFECT OF CHEMISTRY VARIATIONS IN PLATE AND WELD FILLER METAL ON THE CORROSION PERFORMANCE OF NI-CR-MO ALLOYS

    SciTech Connect

    D.V. Fix

    2006-02-07

    The ASTM standard B 575 provides the requirements for the chemical composition of Nickel-Chromium-Molybdenum (Ni-Cr-Mo) alloys such as Alloy 22 (N06022) and Alloy 686 (N06686). The compositions of each element are given in a range. For example, the content of Mo is specified from 12.5 to 14.5 weight percent for Alloy 22 and from 15.0 to 17.0 weight percent for Alloy 686. It was important to determine how the corrosion rate of welded plates of Alloy 22 using Alloy 686 weld filler metal would change if heats of these alloys were prepared using several variations in the composition of the elements even though still in the range specified in B 575. All the material used in this report were especially prepared at Allegheny Ludlum Co. Seven heats of plate were welded with seven heats of wire. Immersion corrosion tests were conducted in a boiling solution of sulfuric acid plus ferric sulfate (ASTM G 28 A) using both as-welded (ASW) coupons and solution heat-treated (SHT) coupons. Results show that the corrosion rate was not affected by the chemistry of the materials in the range of the standards.

  18. IBA analysis and corrosion resistance of TiAlPtN/TiAlN/TiAl multilayer films deposited over a CoCrMo using magnetron sputtering

    NASA Astrophysics Data System (ADS)

    Canto, C. E.; Andrade, E.; de Lucio, O.; Cruz, J.; Solís, C.; Rocha, M. F.; Alemón, B.; Flores, M.; Huegel, J. C.

    2016-03-01

    The corrosion resistance and the elemental profile of multilayer coatings of TiAlPtN/TiAlN/TiAl synthesized by Physical Vapor Deposition (PVD) reactive magnetron sputtering over a CoCrMo alloy substrate in 10 periods of 30 min each were analyzed and compared to those of the substrate alone and to that of a TiAlPtN single layer coating of the same thickness. The objective of the present work was to create multilayers with different amounts of Pt to enhance the corrosion resistance of a biomedical alloy of CoCrMo. Corrosion tests were performed using Simulated Body Fluid (SBF) using potentiodynamic polarization tests at typical body temperature. The elemental composition and thickness of the coatings were evaluated with the combination of two ion beam analysis (IBA) techniques: a Rutherford Backscattering Spectroscopy (RBS) with alpha beam and a Nuclear Reaction Analysis with a deuteron beam.

  19. Comparative Study on the Corrosion Resistance of Fe-Based Amorphous Metal, Borated Stainless Steel and Ni-Cr-Mo-Gd Alloy

    SciTech Connect

    Lian, Tiangan; Day, Daniel; Hailey, Phillip; Choi, Jor-Shan; Farmer, Joseph

    2007-07-01

    Iron-based amorphous alloy Fe{sub 49.7}Cr{sub 17.7}Mn{sub 1.9}Mo{sub 7.4}W{sub 1.6}B{sub 15.2}C{sub 3.8}Si{sub 2.4} was compared to borated stainless steel and Ni-Cr-Mo-Gd alloy on their corrosion resistance in various high-concentration chloride solutions. The melt-spun ribbon of this iron-based amorphous alloy have demonstrated a better corrosion resistance than the bulk borated stainless steel and the bulk Ni-Cr-Mo-Gd alloy, in high-concentration chloride brines at temperatures 90 deg. C or higher. (authors)

  20. Alloying the X40CrMoV5-1 steel surface layer with tungsten carbide by the use of a high power diode laser

    NASA Astrophysics Data System (ADS)

    Dobrzański, L. A.; Bonek, M.; Hajduczek, E.; Klimpel, A.

    2005-07-01

    The paper presents the effect of alloying with tungsten carbide on properties of the X40CrMoV5-1 steel surface layer, using the high power diode laser (HPDL). Selection of laser operating conditions is discussed, as well as thickness of the alloying layer, and their influence on structure and chemical composition of the steel. Analysis of the influence of the process conditions on the thicknesses of the alloyed layer and heat-affected zone is presented.

  1. Tribocorrosive behaviour of commonly used temporomandibular implants in a synovial fluid-like environment: Ti-6Al-4V and CoCrMo

    NASA Astrophysics Data System (ADS)

    Royhman, D.; Yuan, J. C.; Shokuhfar, T.; Takoudis, C.; Sukotjo, C.; Mathew, M. T.

    2013-10-01

    The temporomandibular joint implant metal alloys, Ti6Al4V and CoCrMo, (n = 3/group) were tested under free-potential and potentiostatic conditions using a custom-made tribocorrosion apparatus. Sliding duration (1800 cycles), frequency (1.0 Hz) and load (16 N) mimicked the daily mastication process. Synovial-like fluid (bovine calf serum, pH = 7.6 at 37 °C) was used to simulate the in vivo environment. Changes in friction coefficient were monitored throughout the sliding process. Changes in surface topography, total weight loss and roughness values were calculated using scanning electron microscopy and white-light interferometry. Finally, statistical analyses were performed using paired t-tests to determine significance between regions within each metal type and also independent sample t-tests to determine statistical significance between metal alloy types. Ti6Al4V demonstrated a greater decrease of potential than CoCrMo, a higher weight loss from wear (Kw = 257.8 versus 2.62 µg p < 0.0001), a higher weight loss from corrosion (Kc = 17.44 versus 0.14 µg p < 0.0001) and a higher weight loss from the combined effects of wear and corrosion (Kwc = 275.28 versus 2.76 µg p < 0.0001). White-light interferometry measurements demonstrated a greater difference in surface roughness inside the wear region in Ti6Al4V than CoCrMo after the sliding (Ra = 323.80 versus 70.74 nm p < 0.0001). In conclusion, CoCrMo alloy shows superior anti-corrosive and biomechanical properties.

  2. The effect of neutron irradiation on the properties of AlGaAs/GaAs laser diodes

    NASA Technical Reports Server (NTRS)

    Barnes, C. E.; Heflinger, D.; Reel, R.

    1990-01-01

    The effects of neutron irradiation on several properties of both single and multiple stripe laser diodes have been examined. Prior to fast neutron irradiation, total light output as a function of laser current, threshold current, near-field pattern, far-field pattern, and laser output wavelength spectra were measured at room temperature. These measurements were then repeated at intermittent neutron fluence levels. It was observed that the threshold current increased with neutron fluence for all devices examined. In contrast, neutron irradiation had only an indirect effect on the remainder of the laser diode properties in that the higher currents required for operation after irradiation caused variations in these properties.

  3. Effect of aqueous solution and load on the formation of DLC transfer layer against Co-Cr-Mo for joint prosthesis.

    PubMed

    Guo, Feifei; Zhou, Zhifeng; Hua, Meng; Dong, Guangneng

    2015-09-01

    Diamond-like carbon (DLC) coating exhibits excellent mechanical properties such as high hardness, low friction and wear, which offer a promising solution for the metal-on-metal hip joint implants. In the study, the hydrogen-free DLC coating with the element Cr as the interlay addition was deposited on the surface of the Co-Cr-Mo alloy by a unbalanced magnetron sputtering method. The coating thickness was controlled as 2 µm. Nano-indentation test indicated the hardness was about 13 GPa. DLC coated Co-Cr-Mo alloy disc against un-coated Co-Cr-Mo alloy pin (spherical end SR9.5) comprised the friction pairs in the pin-on-disc tribotest under bovine serum albumin solution (BSA) and physiological saline(PS).The tribological behavior under different BSA concetrations(2-20 mg/ml), and applied load (2-15N) was investigated.DLC transfer layer did not form under BSA solution, even though different BSA concetration and applied load changed. The coefficient of friction(COF) under 6 mg/ml BSA at 10 N was the lowest as 0.10. A higher COF of 0.13 was obtained under 20 mg/ml BSA. The boundary absorption layer of protein is the main factor for the counterparts. However, the continous DLC transfer layer was observed under PS solution, which make a lower COF of 0.08. PMID:25967039

  4. Combustion Synthesis of TiB2-TiC/42CrMo4 Composites with Gradient Joint Prepared in Different High-Gravity Fields

    NASA Astrophysics Data System (ADS)

    Huang, Xuegang; Huang, Jie; Zhao, Zhongmin; Yin, Chun; Zhang, Long; Wu, Junyan

    2015-12-01

    The novel TiB2-TiC/42CrMo4-laminated composite materials were successfully fabricated by combustion synthesis in different high-gravity fields. This ceramic/metal composite material possesses continuously graded composition, and multilevel gradient microstructure, which is composed of TiB2-TiC ceramic substrate, ceramic-based intermediate layer, metal-based intermediate layer, and 42CrMo4 substrate. The ceramic-based intermediate layer is the main gradient transition region in the joint which shows that the TiB2 and TiC grains decrease gradually in size and volume fraction from the ceramic substrate to metal substrate. The experiment reveals that the increased high-gravity field not only leads to the higher combustion temperature and the remarkable thermal explosion mode, but also attributes to the enhanced interdiffusion and convection between the molten steel surface and liquid TiB2-based ceramic. So, the reliable fusion bonding of TiB2-TiC/42CrMo4 composite materials is achieved. Moreover, the phase separation and forced filling effect of high-gravity field is the key to improve the densification and bond performance of the joint. The ceramic/metal joint in the continuous gradient composition and microstructure represents not only the transitional change of Vickers hardness, but also the high shear bond strength of 420 ± 25 MPa.

  5. Development of ferritic steels for fusion reactor applications

    SciTech Connect

    Klueh, R.L.; Maziasz, P.J.; Corwin, W.R.

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

  6. The effect of contact load on CoCrMo wear and the formation and retention of tribofilms

    PubMed Central

    Wimmer, M.A.; Laurent, M.P.; Mathew, M.T.; Nagelli, C.; Liao, Y.; Marks, L.D.; Jacobs, J.J.; Fischer, A.

    2015-01-01

    Tribochemical reactions in a protein lubricated metal-on-metal (MoM) sliding contact may play a significant role for its wear performance. Such reactions lead to the formation of a carbonaceous ‘tribofilm’, which can act as a protective layer against corrosion and wear. The purpose of this study was to determine the effect of contact load on wear and the formation and retention of tribofilms. Wear tests were performed in a custom-made ball-on-flat testing apparatus that incorporated an electrochemical cell. A ceramic ball was used to articulate against low-carbon wrought CoCrMo alloy pins in bovine serum. Using a range of contact loads at a single potentiostatic condition (close to free potential), weight loss and changes in surface properties were evaluated. We determined that wear was influenced by the loading condition. As expected, wear increased with load, but the association between applied load and measured weight loss was not linear. In the intermediate load region, in the range of 32–48 N (~58–80 MPa), there was more than an order of magnitude drop in the wear per unit load, and the wear versus load data suggested an inflexion point at 49 N. Regression analyses yielded a cubic model (R2=0.991; p=0.0002), where the cubic term, which represents the inflexion, was highly significant (p=0.0021). This model is supported by the observations that the minimum in the friction versus load curve is at 52 N and the highest relative increase in polarization resistance occurred at 49 N. Scanning electron microscopy and Raman spectroscopy indicated the absence of a tribofilm for the low and within the contact area of the high load cases. Synergistic interactions of wear and corrosion seem to play an important role. PMID:26085697

  7. Development of microstructure and mechanical properties during annealing of a cold-swaged Co-Cr-Mo alloy rod.

    PubMed

    Mori, Manami; Sato, Nanae; Yamanaka, Kenta; Yoshida, Kazuo; Kuramoto, Koji; Chiba, Akihiko

    2016-12-01

    In this study, we investigated the evolution of the microstructure and mechanical properties during annealing of a cold-swaged Ni-free Co-Cr-Mo alloy for biomedical applications. A Co-28Cr-6Mo-0.14N-0.05C (mass%) alloy rod was processed by cold swaging, with a reduction in area of 27.7%, and then annealed at 1173-1423K for various periods up to 6h. The duplex microstructure of the cold-swaged rod consisted of a face-centered cubic γ-matrix and hexagonal closed-packed ε-martensite developed during cold swaging. This structure transformed nearly completely to the γ-phase after annealing and many annealing twin boundaries were observed as a result of the heat treatment. A small amount of the ε-phase was identified in specimens annealed at 1173K. Growth of the γ-grains occurred with increasing annealing time at temperatures ≥1273K. Interestingly, the grain sizes remained almost unchanged at 1173K and a very fine grain size of approximately 8μm was obtained. The precipitation that occurred during annealing was attributed to the limited grain coarsening during heat treatment. Consequently, the specimens treated at this temperature showed the highest tensile strength and lowest ductility among the specimens prepared. An elongation-to-failure value larger than 30% is sufficient for the proposed applications. The other specimens treated at higher temperatures possessed similar tensile properties and did not show any significant variations with different annealing times. Optimization of the present rod manufacturing process, including cold swaging and interval annealing heat treatment, is discussed. PMID:27500542

  8. Role of gaseous environment and secondary precipitation in microstructural degradation of Cr-Mo steel weldments at high temperatures

    SciTech Connect

    Raman, R.K.S.

    1999-08-01

    This study is an attempt to understand the combined role of variations in oxidizing environment and secondary precipitation, in the microstructurally different regions of a standard Cr-Mo steel weldment, on the intensity of internal oxidation during high-temperature oxidation in air and steam environments. Samples of the weld-metal, heat affected zone (HAZ), and base-metal regions were separated from the weldment of 2.25Cr-1Mo steel and oxidized in the environments of air and steam at 873 K. The oxide scales and underlying subscales were characterized using scanning electron microscopy (SEM), energy-dispersive X-ray (EDX) analysis, and electron probe microanalysis (EPMA). Extensive internal oxidation and oxidation-induced void formation in the subscale zone and grain-boundary cavitation in the neighboring region were found to occur during oxidation in the steam environment. However, the internal oxidation and void formation were much more extensive in the subscale regions of the HAZ than in the subscales of the weld-metal and base-metal regions. As a result, the alloy matrix in the area neighboring the subscale region of the HAZ specimen suffered extensive grain-boundary cavitation. This behavior has been attributed to a rather specific combination and complex interplay of the environment, alloy microstructure, oxidizing temperature, and nature of the resulting external scale in causing and sustaining internal oxidation. The article also discusses the role of internal oxidation-assisted microstructural degradation in deteriorating the service life of components of 2.25 Cr-1Mo steel.

  9. 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.

  10. Model of defect reactions and the influence of clustering in pulse-neutron-irradiated Si

    SciTech Connect

    Myers, S. M.; Cooper, P. J.; Wampler, W. R.

    2008-08-15

    Transient reactions among irradiation defects, dopants, impurities, and carriers in pulse-neutron-irradiated Si were modeled taking into account the clustering of the primal defects in recoil cascades. Continuum equations describing the diffusion, field drift, and reactions of relevant species were numerically solved for a submicrometer spherical volume, within which the starting radial distributions of defects could be varied in accord with the degree of clustering. The radial profiles corresponding to neutron irradiation were chosen through pair-correlation-function analysis of vacancy and interstitial distributions obtained from the binary-collision code MARLOWE, using a spectrum of primary recoil energies computed for a fast-burst fission reactor. Model predictions of transient behavior were compared with a variety of experimental results from irradiated bulk Si, solar cells, and bipolar-junction transistors. The influence of defect clustering during neutron bombardment was further distinguished through contrast with electron irradiation, where the primal point defects are more uniformly dispersed.

  11. 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.

  12. Microstructure of V-4Cr-4Ti following low temperature neutron irradiation

    SciTech Connect

    Rice, P.M.; Snead, L.L.; Alexander, D.J.; Zinkle, S.J.

    1996-12-31

    The V-4Cr-4Ti alloys displays excellent mechanical properties, including a ductile-to-brittle transition temperature (DBTT) below - 200 C in the unirradiated conditions. Samples were fission neutron- irradiated in HFBR to a 0.4 dpa dose at 100-275 C. Mechanical tests showed significant irradiation hardening which increased with irradiation temperature. Charpy impact testing also showed a dramatic increase in DBTT on the order of 100 to 350 C. The mechanical property changes are correlated with preliminary results from TEM analysis of the defect microstructure resulting from the low-dose neutron irradiations. TEM of the irradiated material showed a nearly constant defect density of {approximately}1.6x10{sup 23}m{sup -3}, with an average defect diameter of slightly greater than 3 nm.

  13. Detection of previous neutron irradiation and reprocessing of uranium materials for nuclear forensic purposes.

    PubMed

    Varga, Zsolt; Surányi, Gergely

    2009-04-01

    The paper describes novel analytical methods developed for the detection of previous neutron irradiation and reprocessing of illicit nuclear materials, which is an important characteristic of nuclear materials of unknown origin in nuclear forensics. Alpha spectrometry and inductively coupled plasma sector-field mass spectrometry (ICP-SFMS) using solution nebulization and direct, quasi-non-destructive laser ablation as sample introduction were applied for the measurement of trace-level (232)U, (236)U and plutonium isotopes deriving from previous neutron irradiation of uranium-containing nuclear materials. The measured radionuclides and isotope ratios give important information on the raw material used for fuel production and enable confirm the supposed provenance of illicit nuclear material. PMID:19179085

  14. Nitrogen-promoted formation of graphite-like aggregations in SiC during neutron irradiation

    SciTech Connect

    Wang, P. F.; Ruan, Y. F.; Huang, L.; Zhu, W.

    2012-03-15

    The undoped and nitrogen-doped SiC bulk crystals irradiated with two neutron fluences were investigated by using confocal micro-Raman spectroscopy to analyze the effect of nitrogen impurity on irradiation damage. We found that the nitrogen impurity can promote the segregation of carbon atoms into graphite during heavy neutron irradiation, demonstrated by the presence of typical D and G graphite bands. Further experimental analysis indicated that the graphite-like aggregations uniformly distribute in SiC and possess much inferior thermal stability to crystalline graphite. The nucleation, namely generation of stable sp{sup 2} C=C configuration induced by nitrogen atoms, and growth during neutron irradiation can account for the formation of graphite-like aggregations.

  15. Grain boundary segregation in neutron-irradiated 304 stainless steel studied by atom probe tomography

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Nozawa, Y.; Van Renterghem, W.; Matsukawa, Y.; Hatakeyama, M.; Nagai, Y.; Al Mazouzi, A.; Van Dyck, S.

    2012-06-01

    Radiation-induced segregation (RIS) of solute atoms at a grain boundary (GB) in 304 stainless steel (SS), neutron-irradiated to a dose of 24 dpa at 300 °C in the fuel wrapper plates of a commercial pressurized water reactor, was investigated using laser-assisted atom probe tomography (APT). Ni, Si, and P enrichment and Cr and Fe depletion at the GB were evident. The full-width at half-maximum of the RIS region was ˜3 nm for the concentration profile peaks of Ni and Si. The atomic percentages of Ni, Si, and Cr at the GB were ˜19%, ˜7%, and ˜14%, respectively, in agreement with previously-reported values for neutron-irradiated SS. A high number density of intra-granular Ni-Si rich precipitates formed in the matrix. A precipitate-denuded zone with a width of ˜10 nm appeared on both sides of the GB.

  16. Identification of neutron irradiation induced strain rate sensitivity change using inverse FEM analysis of Charpy test

    NASA Astrophysics Data System (ADS)

    Haušild, Petr; Materna, Aleš; Kytka, Miloš

    2015-04-01

    A simple methodology how to obtain additional information about the mechanical behaviour of neutron-irradiated WWER 440 reactor pressure vessel steel was developed. Using inverse identification, the instrumented Charpy test data records were compared with the finite element computations in order to estimate the strain rate sensitivity of 15Ch2MFA steel irradiated with different neutron fluences. The results are interpreted in terms of activation volume change.

  17. 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.

  18. 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.

  19. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  20. New ferritic steels increase the thermal efficiency of steam turbines

    SciTech Connect

    Mayer, K.H.; Bakker, W.T.

    1996-12-31

    The further development of ferritic high-temperature-resistant 9--11%Cr steels has paved the way for fossil-fired power stations to be operated at turbine steam inlet temperatures of up to around 600 C and high supercritical steam pressures with a distinct improvement in thermal efficiency, a significant contribution towards reducing the environmental impact of SO{sub 2}, NO{sub x} and CO{sub 2} emissions and to a more economical utilization of fossil fuels. Advances in the development of these steels are primarily attributable to joint research projects undertaken by the manufacturers and operators of power stations in Japan (EPDC), in the USA (EPRI) and in Europe (COST 501). The report gives details on the results achieved under EPRI Research Project RP 140 3-15/23 on the creep behavior of modified 9%CrMo cast steel used in the manufacture of steam turbines for coal-fired power plants. The modified 9%CrMo cast steel also offers great benefits as regards improving the useful life and thermal efficiency of existing power plants.

  1. In vitro corrosion testing of PVD coatings applied to a surgical grade Co-Cr-Mo alloy.

    PubMed

    Bolton, J; Hu, X

    2002-06-01

    Toxic effects and biological reaction of metallic corrosion and wear products are an important concern for metal on metal artificial joints. Corrosion tests were conducted to study the susceptibility to pitting and localized corrosion, with three coatings, CrN, TiN and DLC, applied to a wrought high carbon Co-Cr-Mo alloy substrate material. Corrosion testing involved the measurement of potential time transients during immersion in a physiological solution and cyclic polarization of specimen potentials into the transpassive range followed by reversal of the potential to scan in the cathodic direction to regain the rest potential E(rest). Resistance to pitting and localized corrosion was assessed by determining the transpassive breakdown potential E(bd) and if any hysteresis generated during the reverse cyclic scan may have caused crossover with the original anodic scan. Three different surface coating conditions were tested namely: (1) as-coated, (2) polished, and (3) indented to penetrate the coating by diamond pyramid hardness indentor. Results showed that all three coatings produced significant improvements in corrosion resistance compared to performance of the wrought cobalt alloy but that some corrosive attack to both the CrN and TiN coatings occurred and some risk of attack to the cobalt alloy substrate existed due to coating defects or when damage to the coating occurred. TiN coatings were highly effective in preventing corrosion provided they were thick enough to produce complete coverage. Thin TiN coatings displayed some tendency to encourage localized attack of the cobalt alloy at coating defects or where the coating suffered mechanical damage. CrN coatings underwent transpassive breakdown more easily and some degree of pitting at defects within the coating was observed, especially when the CrN coating was polished before the test. No corrosive attack of the cobalt alloy substrate was observed when the CrN coating was mechanically damaged by indentation

  2. Evaluation of an Oxide Layer on NI-CR-MO-W Alloy Using Electrochemical Impedance Spectroscopy and Surface Analysis

    SciTech Connect

    D. Zagidulin; P. Jakupi; J.J. Noel; D.W. Shoesmith

    2006-12-21

    High corrosion resistance under very aggressive conditions is a distinguishing property of Ni-Cr-Mo-W alloys. One such alloy, Alloy 22, is a candidate material for fabrication of the outer layer of high-level nuclear waste (HLNW) packages for the proposed HLNW repository at Yucca Mountain, Nevada, USA. We are using Electrochemical Impedance Spectroscopy (EIS), ex-situ X-Ray Photoelectron Spectroscopy (XPS) and Time of Flight Secondary Ion Mass Spectroscopy (ToF SIMS) to characterize the electrochemical properties and composition of the protective oxide formed on Alloy 22 surfaces. These studies have been conducted at temperatures up to 90 C at potentials from -0.8 V to 0.8 V (vs. Ag/AgCl (sat'd KCl)) in deaerated 5 mol L{sup -1} NaCl solution. Using this combination of techniques, we can correlate the electrical (from EIS) and compositional properties (from XPS, ToF SIMS) of the oxide. At more negative potentials (-0.8 V to -0.4 V) the film exhibits a low charge transfer resistance and high capacitance, indicating the presence of a very defective film with a high concentration of electronic defects. The presence of additional elements in the equivalent circuit, corresponding to water reduction, supports this suggestion. At these potentials, surface analysis techniques show a thin oxide layer with a low concentration of Cr203. Increasing the potential (to between -0.2 and 0.2 V) leads to a major increase in overall interfacial resistance consistent with the formation of an oxide with a small concentration of electronic defects. At the same time, the surface analysis techniques show increases in the film thickness and the Cr{sub 2}O{sub 3} content. A further increase in potential to 0.8 V, in general, leads to a decrease in interfacial resistance throughout the film. When the Cr{sub 2}O{sub 3} barrier layer is degraded, then the higher oxidation states of Mo and W species (MO{sup VI}, W{sup VI}) increase in concentration and are stored in the outer part of the film

  3. Combined effect of rapid nitriding and plastic deformation on the surface strength, toughness and wear resistance of steel 38CrMoAlA

    NASA Astrophysics Data System (ADS)

    Wang, B.; Lv, Z. Q.; Zhou, Z. A.; Sun, S. H.; Huang, X.; Fu, W. T.

    2015-08-01

    The combined treatment of pressurized gas nitriding and cold rolling is proposed as a new approach to rapid preparation of a strong and tough nitrided layer for steel 38CrMoAlA. The microstructural characteristics and properties of the modified surface layer in comparison with those of the conventionally gas nitrided sample have systematically been evaluated. The results show that the hardness and toughness of the nitrided surface layer can be significantly improved by the combined treatment. Especially, the wear resistance of nitrided surface layer under heavy loads was greatly enhanced. It can provide a new approach to rapidly preparing a nitrided layer with high strength and toughness.

  4. The double perovskite oxide Sr2CrMoO(6-δ) as an efficient electrocatalyst for rechargeable lithium air batteries.

    PubMed

    Ma, Zhong; Yuan, Xianxia; Li, Lin; Ma, Zi-Feng

    2014-12-01

    A double perovskite oxide Sr2CrMoO6-δ (SCM), synthesized using the sol-gel and annealing method with the assistance of citric acid and ethylene diamine tetraacetic acid, was investigated for the first time as an efficient catalyst for rechargeable lithium air batteries. The SCM cathode enables higher specific capacity, lower overpotential and a much better cyclability compared to the pure Super P electrode owing to its excellent electrocatalytic activity towards the formation/decomposition of Li2O2. PMID:25325080

  5. Extraction of protactinium-233 and separation from thermal neutron-irradiated thorium-232 using crown ethers

    SciTech Connect

    Jalhoom, Moayyed G.; Mohammed, Dawood A.; Khalaf, Jumah S.

    2008-07-01

    A new method was developed for the extraction and separation of {sup 233}Pa from thermal neutron-irradiated {sup 232}Th. Solutions of Pa{sup 233} were prepared in LiCI-HCl solutions from which appreciable extraction was obtained using dibenzo-18-crown-6 in 1,2-dichloroethane. The effects of cavity size, substitutions on the crown ring, type of the organic solvent, and temperature on extraction are discussed. Very high separation factors were obtained for the pairs {sup 233}Pa/{sup 232}Th (>105), {sup 233}Pa/{sup 233}U (> 1000), and {sup 232}U/{sup 232}Th (>60). (authors)

  6. Cation disorder determined by MAS {sup 27}Al NMR in high dose neutron irradiated spinel

    SciTech Connect

    Cooper, E.A.; Sickafus, K.E.; Hughes, C.D.; Earl, W.L.; Hollenberg, G.W.; Garner, F.A.; Bradt, R.C.

    1995-12-31

    Spinel (MgAl{sub 2}O{sub 4}) single crystals which had been neutron irradiated to high doses (53-250 dpa) were examined using {sup 27}Al magic angle spinning (MAS) nuclear magnetic resonance (NMR). The sensitivity of this procedure to a specific cation (Al) residing in different crystallographic environments allowed one to determine the distribution of the Al between the two cation sites in the spinel structure. The samples were irradiated at two different temperatures (400 and 750{degrees}C) and various doses. These results indicate that the Al was nearly fully disordered over the two lattice sites after irradiation.

  7. Thermal annealing of stabilization products from recoil bromine-82 atoms in neutron-irradiated ammonium perbromate

    SciTech Connect

    Isupov, V.K.; Gavrilov, V.V.

    1987-11-01

    A study has been made on the thermal annealing of stabilization products from recoil bromine-82 atoms in neutron-irradiated ammonium perbromate. Paper and ion-exchange chromatography show that the oxidation of /sup 82/Br/sup -/ to /sup 82/BrO/sub 3//sup -/ in that case occurs only to a small extent, in contrast to alkali-metal perbromates. The effect is ascribed to metastable radiolysis products from the ammonium group. The pyrolysis of ammonium perbromate has also been examined.

  8. The observation of structural defects in neutron-irradiated lithium-doped silicon solar cells

    NASA Technical Reports Server (NTRS)

    Sargent, G. A.

    1971-01-01

    Electron microscopy has been used to observe the distribution and morphology of lattice defects introduced into lithium-doped silicon solar cells by neutron irradiation. Upon etching the surface of the solar cells after irradiation, crater-like defects are observed that are thought to be associated with the space charge region around vacancy clusters. Thermal annealing experiments showed that the crater defects were stable in the temperature range 300 to 1200 K in all of the lithium-doped samples. Some annealing of the crater defects was observed to occur in the undoped cells which were irradiated at the lowest doses.

  9. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    SciTech Connect

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  10. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    SciTech Connect

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-03-17

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10{sup 7}. Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays.

  11. Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

    SciTech Connect

    Masashi Shimada; M. Hara; T. Otsuka; Y. Oya; Y. Hatano

    2014-05-01

    Accurately estimating tritium retention in plasma facing components (PFCs) and minimizing its uncertainty are key safety issues for licensing future fusion power reactors. D-T fusion reactions produce 14.1 MeV neutrons that activate PFCs and create radiation defects throughout the bulk of the material of these components. Recent studies show that tritium migrates and is trapped in bulk (>> 10 µm) tungsten beyond the detection range of nuclear reaction analysis technique [1-2], and thermal desorption spectroscopy (TDS) technique becomes the only established diagnostic that can reveal hydrogen isotope behavior in in bulk (>> 10 µm) tungsten. Radiation damage and its recovery mechanisms in neutron-irradiated tungsten are still poorly understood, and neutron-irradiation data of tungsten is very limited. In this paper, systematic investigations with repeated plasma exposures and thermal desorption are performed to study defect annealing and thermal desorption of deuterium in low dose neutron-irradiated tungsten. Three tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to high flux (ion flux of (0.5-1.0)x1022 m-2s-1 and ion fluence of 1x1026 m-2) deuterium plasma at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy (TDS) was performed with a ramp rate of 10 °C/min up to 900 °C, and the samples were annealed at 900 °C for 0.5 hour. These procedures were repeated three (for 100 and 200 °C samples) and four (for 500 °C sample) times to uncover damage recovery mechanisms and its effects on deuterium behavior. The results show that deuterium retention decreases approximately 90, 75, and 66 % for 100, 200, and 500 °C, respectively after each annealing. When subjected to the same TDS recipe, the desorption temperature shifts from 800 °C to 600 °C after 1st annealing

  12. Microstructural examination of fast-neutron irradiated Li/sub 2/O

    SciTech Connect

    Liu, Y.Y.; Mattas, R.F.; Smith, D.L.; Porter, D.L.

    1984-10-01

    Scanning Electron Microscopy examinations of fast-neutron irradiated Li/sub 2/O at 608 to 625/sup 0/C to 1 and 3 at. % /sup 6/Li burnups have been performed. Of particular interests are the helium bubble morphologies and their relationships to the reported tritium/helium retentions and swelling in Li/sub 2/O. Possible defect trapping of tritium is suggested, along with discussions of two other phenomena (microcracking, swelling) on the performance of an Li/sub 2/O fusion reactor blanket.

  13. Effects of neutron irradiation of ultra-thin HfO{sub 2} films

    SciTech Connect

    Hsu, K.-W.; Bian, S.; Shohet, J. L.; Ren, H.; Agasie, R. J.; Nishi, Y.

    2014-01-20

    Neutron irradiation at low fluence decreases the Pb-type and E′ defect levels in ultra-thin hafnium dioxide films because electrons can fill existing states. These electrons come from electron-hole pairs generated by neutron interactions with silicon and oxygen. Thus, a low fluence of neutrons “anneals” the sample. However, when neutron fluence increases, more neutrons collide with oxygen atoms and cause them to leave the lattice or to transmute into different atoms. This causes the E′ states to increase. As defect-state concentrations increase, leakage currents increase, but since the E′ is much lower than the Pb concentration, this is not a dominant factor.

  14. Boundary lubrication of stainless steel and CoCrMo alloy based on phosphorous and boron compounds in oil-in-water emulsion

    NASA Astrophysics Data System (ADS)

    Yan, Jincan; Zeng, Xiangqiong; Ren, Tianhui; van der Heide, Emile

    2014-10-01

    Emulsion lubrication is widely used in metal forming operations and has potential applications in the biomedical field, yet the emulsion lubrication mechanism is not well understood. This work explores the possibilities of three different oil-in-water (O/W) emulsions containing dibutyl octadecylphosphoramidate (DBOP), 6-octadecyl-1,3,6,2-dioxazaborocan-2-ol calcium salt (ODOC) and 2-(4-dodecylphenoxy)-6-octadecyl-1,3,6,2-dioxazaborocane (DOB) to generate boundary films on stainless steel AISI 316 and CoCrMo alloy surfaces. Experimental results show lower friction values for the emulsions in combination with CoCrMo compared to AISI 316. The different performance of the additives is related to the composition of the adsorption and reaction film on the interacting surfaces, which was shown to be dependent on the active elements and molecular structure of the additives. The friction profile of the emulsions indicates that the emulsion appears to be broken during the rubbing process, then the additives adsorb onto the metal surface to form protecting boundary layers. The XPS analysis shows that for boundary lubrication conditions, the additive molecules in the emulsion first adsorb on the metal surface after the droplet is broken, and then decompose and react with the metal surface during the rubbing process to form stable lubricating films on the rubbed surfaces.

  15. 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.

  16. First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

    NASA Astrophysics Data System (ADS)

    Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

    2011-08-01

    With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

  17. 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.

  18. Long-term strength and allowable stresses of grade 10Kh9MFB and X10CrMoVNb9-1 (T91/P91) chromium heat-resistant steels

    NASA Astrophysics Data System (ADS)

    Skorobogatykh, V. N.; Danyushevskiy, I. A.; Schenkova, I. A.; Prudnikov, D. A.

    2015-04-01

    Currently, grade X10CrMoVNb9-1 (T91, P91) and 10Kh9MFB (10Kh9MFB-Sh) chromium steels are widely applied in equipment manufacturing for thermal power plants in Russia and abroad. Compilation and comparison of tensile, impact, and long-term strength tests results accumulated for many years of investigations of foreign grade X10CrMoVNb9-1, T91, P91, and domestic grade 10Kh9MFB (10Kh9MFB-Sh) steels is carried out. The property identity of metals investigated is established. High strength and plastic properties of steels, from which pipes and other products are made, for operation under creep conditions are confirmed. Design characteristics of long-term strength on the basis of tests with more than one million of hour-samples are determined ( and at temperatures of 500-650°C). The table of recommended allowable stresses for grade 10Kh9MFB, 10Kh9MFB-SH, X10CrMoVNb9-1, T91, and P91 steels is developed. The long-time properties of pipe welded joints of grade 10Kh9MFB+10Kh9MFB, 10Kh9MFB-Sh+10Kh9MFB-Sh, X10CrMoVNb9-1+X10CrMoVNb9-1, P91+P91, T91+T91, 10Kh9MFB (10Kh9MFB-Sh)+X10CrMoVNb9-1(T/P91) steels is researched. The welded joint reduction factor is experimentally determined.

  19. Perspectives for online analysis of raw material by pulsed neutron irradiation

    NASA Astrophysics Data System (ADS)

    Bach, Pierre; Le Tourneur, P.; Poumarede, B.

    1997-02-01

    On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

  20. Mechanical properties of advanced SiC/SiC composites after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Ozawa, K.; Nozawa, T.; Katoh, Y.; Hinoki, T.; Kohyama, A.

    2007-08-01

    The effect of neutron irradiation on tensile properties in advanced 2D-SiC/SiC composites was evaluated. The composites used were composed of a SiC matrix obtained by the forced-flow chemical vapor infiltration (FCVI) process and either Tyranno™-SA Grade-3 or Hi-Nicalon™ Type-S fibers with single-layered PyC coating as the interphase. Neutron irradiation fluence and temperature were 3.1 × 10 25 n/m 2 ( E > 0.1 MeV) and 1.2 × 10 26 n/m 2 at 740-750 °C. Tensile properties were evaluated by cyclic tensile test, and hysteresis loop analysis was applied in order to evaluate interfacial properties. Both composites exhibited excellent irradiation resistance in ultimate and proportional limit tensile stresses. From the hysteresis loop analysis, the level of interfacial sliding stress decreased significantly after irradiation to 1.5 × 10 26 n/m 2 at 750 °C.

  1. Neutron irradiation induced microstructural changes in NBG-18 and IG-110 nuclear graphites

    SciTech Connect

    Karthik, Chinnathambi; Kane, Joshua; Butt, Darryl P.; Windes, William E.; Ubic, Rick

    2015-05-01

    This paper reports the neutron-irradiation-induced effects on the microstructure of NBG-18 and IG-110 nuclear graphites. The high-temperature neutron irradiation at two different irradiation conditions was carried out at the Advanced Test Reactor National User Facility at the Idaho National Laboratory. NBG-18 samples were irradiated to 1.54 dpa and 6.78 dpa at 430 °C and 678 °C respectively. IG-110 samples were irradiated to 1.91 dpa and 6.70 dpa at 451 °C and 674 °C respectively. Bright-field transmission electron microscopy imaging was used to study the changes in different microstructural components such as filler particles, microcracks, binder and quinoline-insoluble (QI) particles. Significant changes have been observed in samples irradiated to about 6.7 dpa. The closing of pre-existing microcracks was observed in both the filler and the binder phases. The binder phase exhibited substantial densification with near complete elimination of the microcracks. The QI particles embedded in the binder phase exhibited a complete microstructural transformation from rosettes to highly crystalline solid spheres. The lattice images indicate the formation of edge dislocations as well as extended line defects bridging the adjacent basal planes. The positive climb of these dislocations has been identified as the main contributor to the irradiation-induced swelling of the graphite lattice.

  2. Neutron irradiation study of Nd-Fe-B permanent magnets made from melt-spun ribbons

    NASA Astrophysics Data System (ADS)

    Brown, R. D.; Cost, J. R.; Meisner, G. P.; Brewer, E. G.

    1988-11-01

    Radiation-induced changes in the magnetization of sintered Nd-Fe-B permanent magnets are known to vary widely among specimens produced by different manufacturers. Samples of Nd-Fe-B MAGNEQUENCH magnets, which are made from melt-spun ribbons, have not been studied and show a much reduced sensitivity to neutron irradiation than do sintered Nd-Fe-B magnets. All melt-spun ribbon-based MAGNEQUENCH magnets, i.e., epoxy-bonded, hot-pressed, and die-upset magnets, show essentially the same slow decrease in magnetic remanence with neutron dose. Measurements of the open-circuit remanence Br/Br 0 at various times during the irradiation show a decay of only 1.5% of the preirradiated value for the MAGNEQUENCH magnets after 1 h of irradiation, or a dose of 1.4×1016 neutrons/cm2, compared to a 4.6% drop in remanence for the best sintered Nd-Fe-B magnet (Sumitomo 30H) with the same irradiation dose. Moreover, after 5.3 h of irradiation, the remanence drops by only 3% for the MAGNEQUENCH magnets. Magnets made from melt-spun ribbons are thus the least sensitive to neutron irradiation so far measured for Nd-Fe-B permanent magnets, but are somewhat more sensitive than samarium-cobalt magnets.

  3. The effect of neutron irradiation dose on vacancy defect accumulation and annealing in pure nickel

    NASA Astrophysics Data System (ADS)

    Druzhkov, A. P.; Arbuzov, V. L.; Perminov, D. A.

    2012-02-01

    In order to investigate the dose dependence of vacancy defect evolution in nickel, specimens of high-purity Ni were neutron-irradiated at ˜330 K in the IVV-2M reactor (Russia) to fluencies in the range of 1 × 10 21-1 × 10 23 n/m 2 ( E > 0.1 MeV) corresponding to displacement dose levels in the range of about 0.0001-0.01 dpa and subsequently stepwise annealed to about 900 K. Ni was characterized both in as-irradiated state as well as after post-irradiation annealing by positron annihilation spectroscopy. The formation of three-dimensional vacancy clusters (3D-VCs) in cascades was observed under neutron irradiation, the concentration of 3D-VCs increases with increasing dose level. 3D-VCs collapse into secondary-type clusters (stacking fault tetrahedra (SFTs), and vacancy loops) during stepwise annealing at 350-450 K. It is shown that the thermal stability of SFTs grow with increasing dose level, probably, it is due to growth of the average SFT size during annealing. The results of annealing experiments on electron-irradiated Ni at 300 K are indicated in the paper, for comparison. We also have briefly discussed the positron response to the SFT-like structures.

  4. Neutron irradiation of polycrystalline yttrium aluminate garnet, magnesium aluminate spinel and α-alumina.

    NASA Astrophysics Data System (ADS)

    Neeft, E. A. C.; Konings, R. J. M.; Bakker, K.; Boshoven, J. G.; Hein, H.; Schram, R. P. C.; van Veen, A.; Conrad, R.

    1999-08-01

    Polycrystalline pellets of yttrium aluminate garnet (Y 3Al 5O 12), magnesium aluminate spinel (MgAl 2O 4) and α-alumina (α-Al 2O 3) have been irradiated in the high flux reactor (HFR) at Petten to a neutron fluence of 1.7 × 10 26 m -2 ( E>0.1 MeV) at a temperature of about 815 K. Volume changes smaller than 1% have been measured for Y 3Al 5O 12 and MgAl 2O 4. Transmission electron microscopy (TEM) results of Y 3Al 5O 12 show no difference between the unirradiated TEM samples and neutron-irradiated samples. For MgAl 2O 4, dislocation loops in some grains are found in the irradiated samples. TEM results of Al 2O 3 show a dense network of dislocation loops after neutron irradiation. The increase in volume is 4.2% for a neutron fluence of 1.7 × 10 26 m -2.

  5. Neutron irradiation effect on site distribution of cations in non-stoichiometric magnesium aluminate spinel

    NASA Astrophysics Data System (ADS)

    Sawabe, Takashi; Yano, Toyohiko

    2008-02-01

    Neutron irradiation effects on cation distribution in non-stoichiometric Mg-Al spinel were examined by ALCHEMI (Atom Location by Channeling Enhanced Microanalysis) method. Parameter n, or non-stoichiometry of MgO · nAl 2O 3 of the specimens, were n = 1.00, 1.01, 1.10, 1.48. These specimens were neutron-irradiated up to a fluence of 2.3 × 10 24 n/m 2 ( E > 0.1 MeV) at 500-530 °C in JMTR. Some specimens contracted by the irradiation and the arrangement of cations became more disorder. The other specimens showed very small swelling by the irradiation and the cation distribution became slightly ordered. The cation distribution of the contracted specimen returned stepwise to the pre-irradiated condition after the annealing at 700 °C. The cation distribution of the slightly swollen specimens did not change after the annealing up to 700 °C. Cation distribution in the T-site was more sensitively influenced by the irradiation.

  6. Investigation of the combined effect of neutron irradiation and electron beam exposure on pure tungsten

    NASA Astrophysics Data System (ADS)

    Van Renterghem, W.; Uytdenhouwen, I.

    2016-08-01

    Pure tungsten samples were neutron irradiated in the BR2 reactor of SCK·CEN to fluences of 1.47 × 1020 n/cm2 and 4.74 × 1020 n/cm2 at 300 °C under Helium atmosphere and exposed to the electron beam of the Judith 1 installation The effect of these treatments on the defect structure was studied with transmission electron microscopy. In the irradiated samples the defect structure in the bulk is compared to the structure at the surface. The neutron irradiation created a large amount of a/2‹111› type dislocation loops forming dislocation rafts. The loop density increased from 8.5 × 1021/m³ to 9 × 1022/m³ with increasing dose, while the loop size decreased from 5.2 nm to 3.5 nm. The electron beam exposure induced significant annealing of the defects and almost all of the dislocation loops were removed. The number of line dislocations in that area increased as a result of the thermal stresses from the thermal shock.

  7. Evaluation of Damage Tolerance of Advanced SiC/SiC Composites after Neutron Irradiation

    NASA Astrophysics Data System (ADS)

    Ozawa, Kazumi; Katoh, Yutai; Nozawa, Takashi; Hinoki, Tatsuya; Snead, Lance L.

    2011-10-01

    Silicon carbide composites (SiC/SiC) are attractive candidate materials for structural and functional components in fusion energy systems. The effect of neutron irradiation on damage tolerance of the nuclear grade SiC/SiC composites (plain woven Hi-Nicalon™ Type-S reinforced CVI matrix composites multilayer interphase and unidirectional Tyranno™-SA3 reinforced NITE matrix with carbon mono-layer interphase) was evaluated by means of miniaturized single-edged notched beam test. No significant changes in crack extension behavior and in the load-loadpoint displacement characteristics such as the peak load and hysteresis loop width were observed after irradiation to 5.9 × 1025 n/m2 (E > 0.1 MeV) at 800°C and to 5.8 × 1025 n/m2 at 1300°C. By applying a global energy balance analysis based on non-linear fracture mechanics, the energy release rate for these composite materials was found to be unchanged by irradiation with a value of 3±2 kJ/m2. This has led to the conclusion that, for these fairly aggressive irradiation conditions, the effect of neutron irradiation on the fracture resistance of these composites appears insignificant.

  8. Results of d+T fast neutron irradiation on advanced tumors of bladder and rectum

    SciTech Connect

    Battermann, J.J.

    1982-12-01

    From November, 1975 to November, 1981, around 400 patients were irradiated with 14 MeV d+T fast neutrons at the Antoni van Leeuwenhoek Hospital in Amsterdam. Special interest was focused on inoperable tumors of bladder and rectum. During the pilot phase of the study 47 patients were treated, mostly via two parallel opposed ports with dosages that ranged from 18 to more than 22 Gy. Although persistent local control was achieved in 23 patients (48%), 14 patients (29%) died of severe complications. By the introduction of a six field technique, the fatal complication rate could be reduced significantly. Since May 1978 patients were randomized in a three arm trial, using two dose levels on the neutron site. The preliminary results of a group of 91 patients show a similar survival in the three treatment arms with a somewhat better local control rate for high dose neutrons. An attempt was made to estimate RBE values for tumor control and normal tissue reactions by comparing the data for neutron irradiation with the data obtained with photons on a similar group of patients. From the values derived it must be concluded that the gain for neutron irradiation on these tumors in the pelvis will be negligible.

  9. Results of d+T fast neutron irradiation on advanced tumors of bladder and rectum

    SciTech Connect

    Battermann, J.J.

    1982-12-01

    From November, 1975 to November, 1981, around 400 patients were irradiated with 14 MeV d+T fast neutrons at the Antoni van Leeuwenhoek Hospital in Amsterdam. Special interest was focused on inoperable tumors of bladder and rectum. During the pilot phase of the study 47 patients were treated, mostly via two parallel opposed ports with dosages that ranged from 18 to more than 22 Gy. Although persistent local control was achieved in 23 patients (48%), 14 patients (29%) died of severe complications. By the introduction of a six field technique, the fatal complication rate could be reduced significantly. Since May 1978 patients were randomized in a three arm trial, using two dose levels on the neutron site. The preliminary results of a group of 91 patients show a similar survival in the three treatment arms with a somewhat better local control rate for high dose neutrons. An attempt was made to estimate RBE values for tumor control and normal tissue reaction by comparing the data for neutron irradiation with the data obtained with photons on a similar group of patients. From the values derived it must be concluded that the gain for neutron irradiation on these tumors in the pelvis will be negligible.

  10. Point defects in 4H-SiC epilayers introduced by neutron irradiation

    NASA Astrophysics Data System (ADS)

    Hazdra, Pavel; Záhlava, Vít; Vobecký, Jan

    2014-05-01

    Electronic properties of radiation damage produced in 4H-SiC by neutron irradiation and its effect on electrical parameters of Junction Barrier Schottky (JBS) diodes were investigated. 4H-SiC N-epilayers, which formed the low-doped N-base of JBS power diodes, were irradiated with 1 MeV neutrons with fluences ranging from 1.3 × 1013 to 4.0 × 1014 cm-2. Radiation defects were then characterized by capacitance deep-level transient spectroscopy, I-V and C-V measurement. Results show that neutron irradiation introduces different point defects giving rise to acceptor levels lying 0.61/0.69, 0.88, 1.03, 1.08 and 1.55 eV below the SiC conduction band edge. Introduction rates of these centers are ranging from 0.64 to 4.0 cm-1. These defects have a negligible effect on blocking and dynamic characteristics of irradiated diodes. However, the acceptor character of introduced deep levels and their fast introduction deteriorate diode's ON-state resistance already at fluences exceeding 1 × 1014 cm-2.

  11. Energy spectra of primary knock-on atoms under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Gilbert, M. R.; Marian, J.; Sublet, J.-Ch.

    2015-12-01

    Materials subjected to neutron irradiation will suffer from a build-up of damage caused by the displacement cascades initiated by nuclear reactions. Previously, the main "measure" of this damage accumulation has been through the displacements per atom (dpa) index, which has known limitations. This paper describes a rigorous methodology to calculate the primary atomic recoil events (often called the primary knock-on atoms or PKAs) that lead to cascade damage events as a function of energy and recoiling species. A new processing code SPECTRA-PKA combines a neutron irradiation spectrum with nuclear recoil data obtained from the latest nuclear data libraries to produce PKA spectra for any material composition. Via examples of fusion relevant materials, it is shown that these PKA spectra can be complex, involving many different recoiling species, potentially differing in both proton and neutron number from the original target nuclei, including high energy recoils of light emitted particles such as α-particles and protons. The variations in PKA spectra as a function of time, neutron field, and material are explored. The application of PKA spectra to the quantification of radiation damage is exemplified using two approaches: the binary collision approximation and stochastic cluster dynamics, and the results from these different models are discussed and compared.

  12. Degradation of mechanical properties of stainless steel cladding due to neutron irradiation and thermal aging

    SciTech Connect

    Haggag, F.M.

    1994-09-01

    Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect following neutron irradiation at 288{degrees}C to a fluence of 5 X 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125{degrees}C) and no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub {kappa}}) much more than did thermal aging alone. However, irradiation slightly decreased the tearing modulus but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimens become available. Also, long-term thermal exposure of the three-wire cladding as well as type 308 stainless steel weld materials at 343{degrees}C is in progress.

  13. 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.

  14. The Effect of Neutron Irradiation on the Fracture Toughness of Graphite

    SciTech Connect

    Burchell, Timothy D; Strizak, Joe P

    2012-01-01

    As part of our irradiated graphite recycle program a small quantity of PCEA grade graphite was irradiated in the High Flux Isotope Reactor (HFIR) at ORNL. The graphite will provide the raw material for future recycle experiments. The geometry of the irradiated graphite allowed us to study the effects of neutron irradiation on the Critical Stress Intensity Factor, KIc, of graphite. The specimens where irradiated in two groups of 6 at an irradiation temperature of 900 C in rabbit capsules to doses of 6.6 and 10.2 DPA, respectively. Following a full suite of pre-and post-irradiation examination, which included dimensions, mass, electrical resistivity, elastic constants, and thermal expansion (to 800 C) the samples were notched and tested to determine their KIc using the newly approved ATSM test method for SENB fracture toughness of graphite. Here we report the irradiation induced changes in the dimensions, elastic constants, resistivity, and coefficient of thermal expansion of PCEA graphite. Moreover, irradiation induced changes in the Critical Stress Intensity Factor, KIc, or fracture toughness, are reported and discussed. Very little work on the effect of neutron irradiation on the fracture toughness of graphite has previously be performed or reported.

  15. Design and characterisation of a new duplex surface system based on S-phase hardening and carbon-based coating for ASTM F1537 Co-Cr-Mo alloy

    NASA Astrophysics Data System (ADS)

    Luo, Xia; Li, Xiaoying

    2014-02-01

    Co-Cr-Mo alloys are one of the most widely used metallic biomaterials for metal-on-metal joint prostheses. However, concerns over increased revision rates mainly due to nano-sized wear debris have been raised. This study was aimed at enhancing the friction, wear and load-bearing properties of Co-Cr-Mo alloys by developing a new duplex surface system combining super hard and wear-resistant S-phase layer with self-lubricating, low-friction carbon-based coating. To this end, ASTM

  16. The effect of microstructural changes on the caustic stress corrosion cracking resistance of a NiCrMoV rotor steel

    NASA Astrophysics Data System (ADS)

    Bandyopadhyay, N.; Briant, C. L.; Hall, E. L.

    1985-07-01

    This paper presents a study of the effects of microstructural changes on the caustic stress corrosion cracking resistance of a NiCrMoV rotor steel. All tests were run in 9 M NaOH at 98 °C and at an electrochemical potential of -400 mVHg/Hgo. Different microstructures were obtained by tempering martensitic samples for different times at 600 °C or by using a slow controlled cool from the austenite to produce a bainitic structure. The results show that heat treatments which produced large, chromiumrich carbides are beneficial. These carbides are preferentially corroded and cause pits to form at the crack tip. We propose that these pits cause crack tip blunting and slow crack propagation. It is further shown that, although changes in microstructure can produce improvements in the susceptibility to stress corrosion cracking, these changes cannot compensate for the detrimental effects of phosphorus segregation to grain boundaries.

  17. Importance of metal-metal interactions through the P-P bonds for the multidimensional electrical properties of MP4 (M = V, Cr, Mo)

    SciTech Connect

    Alvarez, S.; Fontcuberta, J.; Myunghwan Whangbo

    1988-07-27

    The electronic structures of MP/sub 4/ (M = V, Cr, Mo) were examined by performing tight-binding band calculations on VP/sub 4/. The MP/sub 4/ phase has a structure in which metal atom chains are sandwiched between P/sub 4//sup 2/minus// layers. Despite this apparently one-dimensional character, the t/sub 2g/-block bands of VP/sub 4/ are dispersive not only along the chain but also along the interchain and interlayer directions. This study shows that the multidimensional nature of those bands originates from the fact that metal atom chains interact with one another through the P-P bonds of the P/sub 4//sup 2/minus// layers. Their Fermi surface calculations reveal that VP/sub 4/ is a three-dimensional metal, while CrP/sub 4/ and MoP/sub 4/ are semimetals. 19 refs., 5 figs., 2 tabs.

  18. Microstructure Investigation of Directionally Solidified NiAl-Cr(Mo)- xDy ( x = 0, 0.1 wt.%) Hypereutectic Alloys at Different Withdrawal Rates

    NASA Astrophysics Data System (ADS)

    Wang, Lei; Shen, Jun; Shang, Zhao; Zhang, Jian-Fei; Du, Yu-Jun; Fu, Heng-Zhi

    2013-11-01

    The microstructures of directionally solidified Ni-31Al-32Cr-6Mo (at.%)- xDy ( x = 0, 0.1 wt.%) hypereutectic alloys were studied at different withdrawal rates. The results show that the microstructure changes from the planar eutectic to the cellular eutectic and the volume fraction of the primary Cr(Mo) dendrites decreases for the Dy-free alloy with the withdrawal rate varying from 6 μm/s to 30 μm/s. The addition of 0.1 wt.% Dy promotes the planar-to-cellular transition. Moreover, the white Dy-containing phase does not form in the alloy for the planar interface growth (6 μm/s), but it can occur in the boundary of eutectic cells for the cellular interface growth (30 μm/s). A sketchy model of the planar and cellular growth is supposed to interpret it.

  19. Insights into the ultrahigh glass-forming ability of the Fe-Co-Cr-Mo-C-B-Y alloy system from the electronic-structure perspective

    NASA Astrophysics Data System (ADS)

    Lu, Y. Z.; Huang, Y. J.; Wang, G.; Shen, J.

    2012-08-01

    The effect of cobalt on the glass-forming ability (GFA) of Fe-Co-Cr-Mo-C-B-Y bulk metallic glasses has been investigated by using ultraviolet photoelectron spectroscopy and X-ray photoelectron spectroscopy. The alloy containing 7% Co (Co7 alloy) exhibits the highest binding energy and the minimum electronic density of states at the Fermi energy in the valence band spectrum and possesses the largest carbide and metal boride peak intensity in the C 1s and B 1s core-level spectrum. The origin of the ultrahigh GFA for the Co7 alloy has been discussed in terms of the unique electronic structures, which are closely related to the densest atomic packing, the smallest atomic clusters, the minimum electronic density of states at the Fermi energy, and the most numerous transition-metal-carbon and transition-metal-boron bonds.

  20. Characterization of air-formed surface oxide film on a Co-Ni-Cr-Mo alloy (MP35N) and its change in Hanks' solution

    NASA Astrophysics Data System (ADS)

    Nagai, Akiko; Tsutsumi, Yusuke; Suzuki, Yuta; Katayama, Keiichi; Hanawa, Takao; Yamashita, Kimihiro

    2012-05-01

    The air-formed surface oxide films used for stents were characterized to determine their composition and chemical state on a Co-Ni-Cr-Mo alloy. The change of the films in Hanks' solution was used to estimate the reconstruction of the film in the human body. Angle-resolved X-ray photoelectron spectroscopy was used to characterize the composition of the film and substrate, as well as the film's thickness. The surface oxide film on the Co-Ni-Cr-Mo alloy (when mechanically polished) consists of oxide species of cobalt, nickel, chromium, and molybdenum, contains a large amount of OH-, and has a thickness of approximately 2.5 nm. Cations exist in the oxide as Co2+, Ni2+, Cr3+, Mo4+, Mo5+, and Mo6+. Chromium is enriched and cobalt and nickel are depleted in the oxide; however, nickel is enriched and cobalt is depleted in the substrate alloy just under the surface oxide film. Concentration of chromium was low and that of nickel was high at small take-off angles. This indicates that distribution of chromium is greater in the inner layer, but nickel is distributed more in the outer layer of the surface oxide film. During immersion in Hanks' solution, cobalt and nickel dissolved, and the film composition changed to mostly chromium oxide (Cr3+), along with small amounts of cobalt, nickel, and molybdenum oxides, and calcium phosphate containing magnesium, potassium, and carbonate. After immersion in Hanks' solution, the thickness of the surface layer containing calcium phosphate increased to more than 4 nm, while the amount of OH- increased. The amount of cobalt and nickel in the surface oxide film and in the substrate alloy just below the oxide decreased during immersion.

  1. Electrochemical control of cell death by reduction-induced intrinsic apoptosis and oxidation-induced necrosis on CoCrMo alloy in vitro.

    PubMed

    Haeri, Morteza; Wӧllert, Torsten; Langford, George M; Gilbert, Jeremy L

    2012-09-01

    Electrochemical voltage shifts in metallic biomedical implants occur in-vivo due to a number of processes including mechanically assisted corrosion. These excursions may compromise the biocompatibility of metallic implants. Voltages can also be controlled to modulate cell function and fate. The in vitro effect of static voltages on the behavior of MC3T3-E1 pre-osteoblasts cultured on CoCrMo alloy (ASTM-1537) was studied to determine the range of cell viability and mode of cell death beyond the viable range. Cell viability and morphology, changes in actin cytoskeleton, adhesion complexes and nucleus, and mode of cell death (necrosis, or intrinsic or extrinsic apoptosis) were characterized at different voltages ranging from -1000 to +500 mV (Ag/AgCl). Moreover, electrochemical currents and metal ion concentrations at each voltage were measured and related to the observed responses. Results show that cathodic and anodic voltages outside the voltage viability range (-400 < V < +500) lead to primarily intrinsic apoptotic and necrotic cell death, respectively. Cell death is associated with cathodic current densities of 0.1 μA cm(-2) and anodic current densities of 10 μA cm(-2). Significant increase in metallic ions (Co, Cr, Ni, Mo) was seen at +500 mV, and -1000 mV (Cr only) compared to open circuit potential. The number and total projected area of adhesion complexes was also lower on the polarized alloy (p < 0.05). These results show that reduction reactions on CoCrMo alloys leads to apoptosis of cells on the surface and may be a relevant mode of cell death for metallic implants in-vivo. PMID:22704843

  2. Tritium Retention and Permeation in Ion- and Neutron-Irradiated Tungsten under US-Japan PHENIX Collaboration

    NASA Astrophysics Data System (ADS)

    Shimada, Masashi; Taylor, Chase N.; Kolasinski, Robert D.; Buchenauer, Dean A.; Chikada, Takumi; Oya, Yasuhisa; Hatano, Yuji

    2015-11-01

    A critical challenge for long-term operation of ITER and beyond to a FNSF, a DEMO and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to intense heat and neutral/ion particle fluxes under the extreme fusion nuclear environment, while minimizing in-vessel inventories and ex-vessel permeation of tritium. Recent work at Tritium Plasma Experiment demonstrated that tritium diffuses in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. US-Japan PHENIX collaboration (2013-2019) investigates irradiation response on tritium behavior in tungsten, and performs one-of-a-kind neutron-irradiation with Gd thermal neutron shield at High Flux Isotope Reactor, ORNL. This presentation describes the challenge in elucidating tritium behavior in neutron-irradiated PFCs, the PHENIX plans for neutron-irradiation and post irradiation examination, and the recent findings on tritium retention and permeation in 14MeV neutron-irradiated and Fe ion irradiated tungsten. This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.

  3. Irradiation creep of low-activation ferritic steels in FFTF/MOTA*1

    NASA Astrophysics Data System (ADS)

    Kohyama, A.; Kohno, Y.; Asakura, K.; Yoshino, M.; Namba, C.; Eiholzer, C. R.

    1994-09-01

    Irradiation creep behavior of low-activation steels, developed as structural materials for fusion reactors (JLF series steels), was investigated to obtain a fundamental understanding of these alloys under fast neutron irradiation in FFTF. (2.25-8)Cr(1-2)W bainitic steels and 12Cr-2W ferritic steels showed superior creep resistance to type-316 stainless steels under fast neutron irradiation up to 520°C. At temperatures below 460°C the creep strain increased with increasing Cr content up to 7 Cr, and further increments of Cr content up to 12% reduced the creep strain. At temperatures between 460 and 600°C, 7-8 Cr ferritic steels showed the largest creep strain. Swelling-enhanced creep, near the peak swelling temperature of 410°C, was also observed. The 9Cr-2W ferritic steel JLF-1 presented excellent properties, suggesting it as a leading candidate alloy for structural components of fusion reactors.

  4. Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

    SciTech Connect

    Chen, Y.; Alexandreanu, B.; Natesan, K.

    2014-11-01

    Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3 were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.

  5. Antiradiation Vaccine: Technology Development Of Prophylaxis, Prevention And Treatment Of Biological Consequences And Complications After Neutron Irradiation.

    NASA Astrophysics Data System (ADS)

    Popov, Dmitri; Maliev, Slava; Jones, Jeffrey

    Introduction: Neutrons irradiation produce a unique biological effectiveness compare to different types of radiation because their ability to create a denser trail of ionized atoms in biological living tissues[Straume 1982; Latif et al.2010; Katz 1978; Bogatyrev 1982]. The efficacy of an Anti-Radiation Vaccine for the prophylaxis, prevention and therapy of acute radiation pathology was studied in a neutron exposure facility. The biological effects of fast neutrons include damage of central nervous system and cardiovascular system with development of Acute Cerebrovascular and Cardiovascular forms of acute radiation pathology. After irradiation by high doses of fast neutron, formation of neurotoxins, hematotoxins,cytotoxins forming from cell's or tissue structures. High doses of Neutron Irradiation generate general and specific toxicity, inflammation reactions. Current Acute Medical Management and Methods of Radiation Protection are not effective against moderate and high doses of neutron irradiation. Our experiments demonstrate that Antiradiation Vaccine is the most effective radioprotectant against high doses of neutron-radiation. Radiation Toxins(biological substances with radio-mimetic properties) isolated from central lymph of gamma-irradiated animals could be working substance with specific antigenic properties for vaccination against neutron irradiation. Methods: Antiradiation Vaccine preparation standard - mixture of a toxoid form of Radiation Toxins - include Cerebrovascular RT Neurotoxin, Cardiovascular RT Neurotoxin, Gastrointestinal RT Neurotoxin, Hematopoietic RT Hematotoxin. Radiation Toxins were isolated from the central lymph of gamma-irradiated animals with different forms of Acute Radiation Syndromes - Cerebrovascular, Cardiovascular, Gastrointestinal, Hematopoietic forms. Devices for Y-radiation were "Panorama","Puma". Neutron exposure was accomplished at the Department of Research Institute of Nuclear Physics, Dubna, Russia. The neutrons

  6. 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.

  7. Quantum transport in neutron-irradiated modulation-doped heterojunctions. I. Fast neutrons

    SciTech Connect

    Jin, W.; Zhou, J.; Huang, Y.; Cai, L.

    1988-12-15

    We have investigated the characteristics of low-temperature quantum transport in Al/sub x/Ga/sub 1-//sub x/As/GaAs modulation-doped heterojunctions irradiated by fast neutrons of about 14 MeV energy. The concentration and the mobility of the two-dimensional electron gas (2D EG) under low magnetic fields decrease with increase in the concentrations of scatterers, such as ionized impurities, lattice defects, and interface roughness. On the other hand, under strong magnetic fields, the Hall plateau broadening associated with the Landau localized states, and the Shubnikov--de Hass (SdH) oscillation enhancement associated with the Landau extended states, increase markedly after fast-neutron irradiation.

  8. Hardness of Carburized Surfaces in 316LN Stainless Steel after Low Temperature Neutron Irradiation

    SciTech Connect

    Byun, TS

    2005-01-31

    A proprietary surface carburization treatment is being considered to minimize possible cavitation pitting of the inner surfaces of the stainless steel target vessel of the SNS. The treatment gives a large supersaturation of carbon in the surface layers and causes substantial hardening of the surface. To answer the question of whether such a hardened layer will remain hard and stable during neutron irradiation, specimens of the candidate materials were irradiated in the High Flux Isotope Reactor (HFIR) to an atomic displacement level of 1 dpa. Considerable radiation hardening occurred in annealed 316LN stainless steel and 20% cold rolled 316LN stainless steel, and lesser radiation hardening in Kolsterised layers on these materials. These observations coupled with optical microscopy examinations indicate that the carbon-supersaturated layers did not suffer radiation-induced decomposition and softening.

  9. 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.

  10. Effects of silicon, carbon and molybdenum additions on IASCC of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Nakano, J.; Miwa, Y.; Kohya, T.; Tsukada, T.

    2004-08-01

    To study the effects of minor elements on irradiation assisted stress corrosion cracking (IASCC), high purity type 304 and 316 stainless steels (SSs) were fabricated and minor elements, Si or C were added. After neutron irradiation to 3.5 × 10 25 n/m 2 ( E>1 MeV), slow strain rate tests (SSRTs) of irradiated specimens were conducted in oxygenated high purity water at 561 K. Specimen fractured surfaces were examined using a scanning electron microscope (SEM) after the SSRTs. The fraction of intergranular stress corrosion cracking (IGSCC) on the fractured surface after the SSRTs increased with neutron fluence. In high purity SS with added C, the fraction of IGSCC was the smallest in the all SSs, although irradiation hardening level was the largest of all the SSs. Addition of C suppressed the susceptibility to IGSCC.

  11. High dose effects in neutron irradiated face-centered cubic metals

    SciTech Connect

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

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors.

  12. Additive analysis of nano silicon under the influence of neutron irradiation

    NASA Astrophysics Data System (ADS)

    Garibli, Aydan; Huseynov, Elchin; Garibov, Adil; Mehdiyeva, Ravan

    2016-04-01

    Nano silicon with 80m2g‑1 specific surface area, 100 nm size and 0.08 g/cm3 density has been irradiated continuously with neutrons (2 × 1013n ṡcm‑2s‑1) up to 20 h at various periods in TRIGA Mark II type research reactor. After the neutron irradiation, cooling time of the samples is taken approximately 360 h. It is found that the initial radioactivity of the irradiated samples changes within 0.1 kBq-3.1 GBq range. Definition of elements’ concentration is determined based on the activities formed in the relevant energy range. After the irradiation, the result of activity analysis carried out the element content of 1% mixture existing in nano Si which has been defined with radionuclides of the relevant element. Moreover, from activities of mixed radioisotopes, their amounts in percentage has been determined.

  13. National Low-Temperature Neutron Irradiation Facility (NLTNIF). The status of development

    SciTech Connect

    Coltman, R.R. Jr.; Kerchner, H.R.; Klabunde, C.E.; Young, F.W. Jr.

    1985-12-01

    In May 1983, the Department of Energy authorized the establishment of a National Low-Temperature Neutron Irradiation Facility (NLTNIF) at ORNL's Bulk Shielding Reactor (BSR). The NLTNIF, which will be available for qualified experiments at no cost to users, will provide a combination of high radiation intensities and special environmental and testing conditions that have not been previously available in the US. Since the DOE authorization, work has proceeded on the design and construction of the new facility without interruption. This report describes the present status of the development of the NLTNIF and the anticipated schedule for completion and performance testing. There is a table of the major specifications and capabilities and a schematic layout of the irradiation cryostate for design and dimensioning of test and experiment assemblies.

  14. Micromechanisms of Twin Nucleation in TiAl: Effects of Neutron Irradiation

    SciTech Connect

    Hishinuma, A.; Yoo, M.H.

    1999-01-28

    The so-called radiation-induced ductility (RID) reported in neutron-irradiated 47at%Al alloys is attributed to the formation of effective twin embryos in the presence of interstitial-type Frank loops in {gamma}-TiAl and the subsequent nucleation and growth of microtwins during post-irradiation tensile deformation. The stability of large faulted Frank loops is explained in terms of the repulsive interaction between Shockley and Frank partials. Interaction of only six ordinary slip dislocations with a Frank loop can facilitate a pole mechanism for twin formation to work. The relative ease of heterogeneous twin nucleation is the reason for the RID and the lack of changes in yield strength and work hardening.

  15. Temperature dependence of the deformation behavior of 316 stainless steel after low temperature neutron irradiation

    SciTech Connect

    Pawel-Robertson, J.E.; Rowcliffe, A.F.; Grossbeck, M.L.

    1996-10-01

    The effects of low temperature neutron irradiation on the tensile behavior of 316 stainless steel have been investigated. A single heat of solution annealed 316 was irradiated to 7 and 18 dpa at 60, 200, 330, and 400{degrees}C. The tensile properties as a function of dose and as a function of temperature were examined. Large changes in yield strength, deformation mode, strain to necking, and strain hardening capacity were seen in this irradiation experiment. The magnitudes of the changes are dependent on both irradiation temperature and neutron dose. Irradiation can more than triple the yield strength over the unirradiated value and decrease the strain to necking (STN) to less than 0.5% under certain conditions. A maximum increase in yield strength and a minimum in the STN occur after irradiation at 330{degrees}C but the failure mode remains ductile.

  16. The effects of neutron irradiation on fracture toughness of austenitic stainless steels.

    SciTech Connect

    Chopra, O. K.; Gruber, E. E.; Shack, W. J.

    1999-05-21

    Austenitic stainless steels are used extensively as structural alloys in reactor pressure vessel internal components because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. This paper presents results of fracture toughness J-R curve tests on four heats of Type 304 stainless steel that were irradiated to fluence levels of {approx}0.3 and 0.9 x 10{sup 21} n cm{sup {minus}2} (E >1 MeV) at {approx}288 C in a helium environment in the Halden heavy water boiling reactor. The tests were performed on 1/4-T compact tension specimens in air at 288 C; crack extensions were determined by both DC potential and elastic unloading compliance techniques.

  17. 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.

  18. Defects in SiO 2 crystals after neutron irradiations at 20 K and 360 K

    NASA Astrophysics Data System (ADS)

    Nakagawa, M.; Okada, M.; Kawabata, Y.; Atobe, K.; Itoh, H.; Nakanishi, S.

    1994-06-01

    The synthetic silicon dioxide (SiO2), cut parallel (x-plate) or perpendicular (z-plate) to c-axis, are irradiated by reactor neutrons at 360 K (2.8 × 1018 n/cm2) or at 20 K (8.0 × 1016 n/cm2). After neutron irradiation at 360 K, the main absorption peak can be observed at 212 nm (5.84 eV) for z-plate and 217 nm (5.71 eV) for x-plate. After irradiation at 20 K a new band at 250 nm (4.96 eV) can be observed in addition to the band at about 220 nm. The 250 nm band having FWHM ∼ 0.44 eV disappears at 300-340 K. Thermoluminescences are also observed between 80 to 400 K; which show some difference between x-plate and z-plate.

  19. Neutron-irradiation effect on the mechanical properties of alumina fiber

    NASA Astrophysics Data System (ADS)

    Sakuma, Yoichi; Iwanaga, Katsusuke; Tsujimoto, Tadashi; Yoshimoto, Takaaki; Okada, Moritami; Miyata, Kiyomi; Iwanaga, Hiroshi

    1998-04-01

    This paper describes the neutron irradiation effects on the deterioration of alumina fiber (made by Mitsui Mining Material, Almax), a typical electrical insulation material. The material was irradiated at the research reactor at Kyoto University Research Reactor Institute with a maximum fluence of 5.6×10 23 n/m 2 (energy: E>0.1 MeV). Tensile strength and tensile modulus of the specimen scarcely changed. Observation with a scanning electron microscope (SEM) and a transmission electron microscope (TEM) did not indicate any changes in crystal or pore structure. However, the Weibull coefficient of tensile strength decreased as the irradiation dose increased. This suggests an increase in the defect size distribution.

  20. 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.

  1. Fast neutron irradiation effects on magnetization relaxation in YBCO single crystals

    SciTech Connect

    Lensink, J.G.; Griessen, R. . Faculty of Physics and Astronomy); Wiesinger, H.P.; Sauerzopf, F.M.; Weber, H.W. ); Crabtree, G.W. )

    1991-07-01

    A high-quality YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} single crystal has been investigated by torque magnetometry prior to and following fast neutron irradiation to a fluence of 2{times}10{sup 21} m{sup {minus}2} (E > 0.1 MeV). In addition to large enhancements of the critical current densities, which have been observed in similar form previously by Sauerzopf et al, we find a dramatic change in the relaxation behavior following irradiation. At low temperatures ({le} 50 k) the relaxation rates are lowered by factors up to 4 in the irradiated state in a magnetic field of 1.5 T. At higher temperatures, on the other hand, they are enhanced compared to the unirradiated state. Both before and after irradiation, the magnetization relaxation follows a logarithmic time dependence, which we ascribe to thermally activated flux motion.

  2. 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.

  3. Separation and Quantification of Chemically Diverse Analytes in Neutron Irradiated Fissile Materials

    SciTech Connect

    Douglas, Matthew; Friese, Judah I.; Greenwood, Lawrence R.; Farmer, Orville T.; Thomas, Linda MP; Maiti, Tapas C.; Finn, Erin C.; Garofoli, Stephanie J.; Gassman, Paul L.; Huff, Morgan M.; Schulte, Shannon M.; Smith, Steven C.; Thomas, Kathie K.; Bachelor, Paula P.

    2009-10-01

    Quantitative measurement of fission and activation products resulting from neutron irradiation of fissile materials is of interest for applications in environmental monitoring, nuclear waste management, and national security. To overcome mass and spectral interferences, and the relative small quantities of some target analytes, an extensive series of chemical separations is necessary. Based on established separations processes involving co-precipitation, solvent extraction, and ion-exchange and extraction chromatography, we have been evaluating and optimizing a proposed sequence of separation steps to allow for the timely quantification of analytes of interest. For simplicity, much of the chemical separation development work has been performed using stable elements as surrogates for the radioactive material. We have recently evaluated the optimized procedures using an irradiated sample to examine the adequacy of separations for measurement of desired analytes by gamma spectrometry. Here we present the results of this evaluation and describe the radiochemical separations utilized.

  4. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    SciTech Connect

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Wirth, Brian D; Snead, Lance Lewis

    2016-01-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

  5. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Snead, Lance L.; Wirth, Brian D.

    2016-03-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (∼90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S-W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage, providing insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

  6. Generation of peanut mutants by fast neutron irradiation combined with in vitro culture

    PubMed Central

    Wang, Jing-Shan; Sui, Jiong-Ming; Xie, Yong-Dun; Guo, Hui-Jun; Qiao, Li-Xian; Zhao, Li-Lan; Yu, Shan-Lin; Liu, Lu-Xiang

    2015-01-01

    Induced mutations have played an important role in the development of new plant varieties. In this study, we investigated the effects of fast neutron irradiation on somatic embryogenesis combined with plant regeneration in embryonic leaflet culture to develop new peanut (Arachis hypogaea L.) germplasm for breeding. The dry seeds of the elite cultivar Luhua 11 were irradiated with fast neutrons at dosages of 9.7, 14.0 and 18.0 Gy. The embryonic leaflets were separated and incubated in a medium with 10.0-mg/l 2,4-D to induce somatic embryogenesis. Next, they were incubated in a medium with 4.0-mg/l BAP for plant regeneration. As the irradiation dosage increased, the frequency of both somatic embryo formation and plantlet regeneration decreased. The regenerated plantlets were grafted onto rootstocks and were transplanted into the field. Later, the mature seeds of the regenerated plants were harvested. The M2 generation plants from most of the regenerated cultivars exhibited variations and segregation in vigor, plant height, branch and pod number, pod size, and pod shape. To determine whether the phenotypes were associated with genomic modification, we compared the DNA polymorphisms between the wild-type plants and 19 M3-generation individuals from different regenerated plants. We used 20 pairs of simple sequence repeat (SSR) primers and detected polymorphisms between most of the mutants and the wild-type plants (Luhua 11). Our results indicate that using a combination of fast neutron irradiation and tissue culture is an effective approach for creating new peanut germplasm. PMID:25653418

  7. Generation of peanut mutants by fast neutron irradiation combined with in vitro culture.

    PubMed

    Wang, Jing-Shan; Sui, Jiong-Ming; Xie, Yong-Dun; Guo, Hui-Jun; Qiao, Li-Xian; Zhao, Li-Lan; Yu, Shan-Lin; Liu, Lu-Xiang

    2015-05-01

    Induced mutations have played an important role in the development of new plant varieties. In this study, we investigated the effects of fast neutron irradiation on somatic embryogenesis combined with plant regeneration in embryonic leaflet culture to develop new peanut (Arachis hypogaea L.) germplasm for breeding. The dry seeds of the elite cultivar Luhua 11 were irradiated with fast neutrons at dosages of 9.7, 14.0 and 18.0 Gy. The embryonic leaflets were separated and incubated in a medium with 10.0-mg/l 2,4-D to induce somatic embryogenesis. Next, they were incubated in a medium with 4.0-mg/l BAP for plant regeneration. As the irradiation dosage increased, the frequency of both somatic embryo formation and plantlet regeneration decreased. The regenerated plantlets were grafted onto rootstocks and were transplanted into the field. Later, the mature seeds of the regenerated plants were harvested. The M2 generation plants from most of the regenerated cultivars exhibited variations and segregation in vigor, plant height, branch and pod number, pod size, and pod shape. To determine whether the phenotypes were associated with genomic modification, we compared the DNA polymorphisms between the wild-type plants and 19 M3-generation individuals from different regenerated plants. We used 20 pairs of simple sequence repeat (SSR) primers and detected polymorphisms between most of the mutants and the wild-type plants (Luhua 11). Our results indicate that using a combination of fast neutron irradiation and tissue culture is an effective approach for creating new peanut germplasm. PMID:25653418

  8. Effect of Neutron Irradiation on Properties of Pb(Mg(1/3)Nb(2/3))O3-PbTiO3.

    PubMed

    Kim, Yong-Il; Choi, Namkyoung; Kim, Geunwoo; Lee, Yun-Hee; Baek, Kwang-Sae; Kim, Ki-Bok

    2015-11-01

    The effect of neutron irradiation on the electrical and piezoelectric properties of a PMN-PT [(Pb(Mg(1/3)Nb(2/3))O3-PbTiO3)] single crystal such as permittivity, electrical impedance and piezoelectric constant d33 has been investigated at 1 kHz. The changes of d33 and permittivity depending on the dose of neutron irradiation for all samples of PMN-PT single crystal were found. In all samples, the permittivity, and piezoelectric constant d33 decreased with the increase of irradiation dose. Changes of XRD patterns depending on the dose of neutron irradiation for all samples were found. From the results of XRDs for analyzing the formation of the PMN-PT single crystals in single phase, the neutron irradiation will affect the crystallinity of PMN-PT single crystals. PMID:26726526

  9. Simultaneous impact of neutron irradiation and sputtering on the surface structure of self–damaged ITER–grade tungsten

    SciTech Connect

    Belyaeva, A. I. Savchenko, A. A.; Galuza, A. A.; Kolenov, I. V.

    2014-07-15

    Simultaneous effects of neutron irradiation and long–term sputtering on the surface relief of ITER–grade tungsten were studied. The effects of neutron–induced displacement damage have been simulated by irradiation of tungsten target with W{sup 6+} ions of 20 MeV energy. Ar{sup +} ions with energy 600 eV were used as imitation of charge exchange atoms in ITER. The surface relief was studied after each sputtering act. The singularity in the WJ–IG surface relief was ascertained experimentally at the first time, which determines the law of roughness extension under sputtering. As follows from the experimental data, the neutron irradiation has not to make a decisive additional contribution in the processes developing under impact of charge exchange atoms only.

  10. Effect of neutron irradiation on mechanical properties of Cu/SS joints after single and multiple HIP cycles

    NASA Astrophysics Data System (ADS)

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

    2000-12-01

    The present design of the ITER plasma facing components consists of a copper alloy heat sink layer between plasma facing materials and stainless steel structure. The main option for manufacturing these components is hot isostatic pressing (HIP) method and several HIP thermal cycles are foreseen for manufacturing of the complete blanket module. Mechanical characterisation of HIP joints between dissimilar metals is a complicated issue, where information on mechanical properties of base alloys, metallurgy of the HIP joints and mechanical testing methods will be required. The tensile and three point bend tests produced different fracture modes, depending on test temperature, applied HIP thermal cycles and neutron irradiation. The fracture mode was either ductile fracture of copper alloy or joint interface fracture. The mechanical properties of the HIP joint specimens were dominated by strength mismatch of the base alloys which was affected by HIP thermal cycles and neutron irradiation.

  11. gamma. -ray and neutron irradiation characteristics of pure silica core single mode fiber and its life time estimation

    SciTech Connect

    Chigusa, Y.; Watanabe, M.; Kyoto, M.; Ooe, M.; Matsubara, T.; Okamoto, S.; Yamamoto, T.; Iida, T.; Sumita, K.

    1988-02-01

    The investigation of the induced loss for a single mode (SM) optical fiber under ..gamma..-ray irradiation and neutron irradiation are described and the estimation method for induced loss with low dose rate and long-term ..gamma..-ray irradiation is proposed. The induced loss of pure silica core SM fiber was estimated to be 50 times lower than that of germanium containing silica core SM fiber after irradiation with 1 R/Hr for 25 years.

  12. The comparison of microstructure and nanocluster evolution in proton and neutron irradiated Fe-9%Cr ODS steel to 3 dpa at 500 °C

    NASA Astrophysics Data System (ADS)

    Swenson, M. J.; Wharry, J. P.

    2015-12-01

    A model Fe-9%Cr oxide dispersion strengthened (ODS) steel was irradiated with protons or neutrons to a dose of 3 displacements per atom (dpa) at a temperature of 500 °C, enabling a direct comparison of ion to neutron irradiation effects at otherwise fixed irradiation conditions. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography including cluster analysis. Both proton and neutron irradiations produced a comparable void and dislocation loop microstructure. However, the irradiation response of the Ti-Y-O oxide nanoclusters varied. Oxides remained stable under proton irradiation, but exhibited dissolution and an increase in Y:Ti composition ratio under neutron irradiation. Both proton and neutron irradiation also induced varying extents of Si, Ni, and Mn clustering at existing oxide nanoclusters. Protons are able to reproduce the void and loop microstructure of neutron irradiation carried out to the same dose and temperature. However, since nanocluster evolution is controlled by both diffusion and ballistic impacts, protons are rendered unable to reproduce the nanocluster evolution of neutron irradiation at the same dose and temperature.

  13. PGNAA system preliminary design and measurement of In-Hospital Neutron Irradiator for boron concentration measurement.

    PubMed

    Zhang, Zizhu; Chong, Yizheng; Chen, Xinru; Jin, Congjun; Yang, Lijun; Liu, Tong

    2015-12-01

    A prompt gamma neutron activation analysis (PGNAA) system has been recently developed at the 30-kW research reactor In-Hospital Neutron Irradiator (IHNI) in Beijing. Neutrons from the specially designed thermal neutron beam were used. The thermal flux of this beam is 3.08×10(6) cm(-2) s(-1) at a full reactor power of 30 kW. The PGNAA system consists of an n-type high-purity germanium (HPGe) detector of 40% efficiency, a digital spectrometer, and a shielding part. For both the detector shielding part and the neutron beam shielding part, the inner layer is composed of (6)Li2CO3 powder and the outer layer lead. The boron-10 sensitivity of the PGNAA system is approximately 2.5 cps/ppm. Two calibration curves were produced for the 1-10 ppm and 10-50 ppm samples. The measurement results of the control samples were in accordance with the inductively coupled plasma atomic emission spectroscopy (ICP-AES) results. PMID:26242556

  14. 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.

  15. Prenatal exposure to gamma/neutron irradiation: Sensorimotor alterations and paradoxical effects on learning

    SciTech Connect

    Di Cicco, D.; Antal, S.; Ammassari-Teule, M. )

    1991-01-01

    The effects of prenatal exposure on gamma/neutron radiations (0.5 Gy at about the 18th day of fetal life) were studied in a hybrid strain of mice (DBA/Cne males x C57BL/Cne females). During ontogeny, measurements of sensorimotor reflexes revealed in prenatally irradiated mice (1) a delay in sensorial development, (2) deficits in tests involving body motor control, and (3) a reduction of both motility and locomotor activity scores. In adulthood, the behaviour of prenatally irradiated and control mice was examined in the open field test and in reactivity to novelty. Moreover, their learning performance was compared in several situations. The results show that, in the open field test, only rearings were more frequent in irradiated mice. In the presence of a novel object, significant sex x treatment interactions were observed since ambulation and leaning against the novel object increased in irradiated females but decreased in irradiated males. Finally, when submitted to different learning tasks, irradiated mice were impaired in the radial maze, but paradoxically exhibited higher avoidance scores than control mice, possibly because of their low pain thresholds. Taken together, these observations indicate that late prenatal gamma/neutron irradiation induces long lasting alterations at the sensorimotor level which, in turn, can influence learning abilities of adult mice.

  16. 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.

  17. 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.

  18. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    NASA Astrophysics Data System (ADS)

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2015-01-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50-70 °C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5 × 1025 m-2 to reach the total ion fluence of 1 × 1026 m-2 in order to investigate the near-surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate the irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near-surface (<5 µm depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near-surface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.3 dpa) at 500 °C cases even in the relatively low ion fluence of 1026 m-2.

  19. 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.

  20. 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.

  1. 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

  2. Filtered fast neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    NASA Astrophysics Data System (ADS)

    Jang, S. Y.; Kim, C. H.; Reece, W. D.; Braby, L. A.

    2004-09-01

    A heavily filtered fast neutron irradiation system (FNIS) was developed for a variety of applications, including the study of long-term health effects of fast neutrons by evaluating the biological mechanisms of damage in cultured cells and living animals such as rats or mice. This irradiation system includes an exposure cave made with a lead-bismuth alloy, a cave positioning system, a gamma and neutron monitoring system, a sample transfer system, and interchangeable filters. This system was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). For a realistic modeling of the NSCR, the irradiation cell, and the FNIS, this study used the Monte Carlo N-Particle (MCNP) code and a set of high-temperature ENDF/B-VI continuous neutron cross-section data. Sensitivity analysis was performed to find the characteristics of the FNIS as a function of the thickness of the lead-bismuth alloy. A paired ion chamber system was constructed with a tissue-equivalent plastic (A-150) and propane gas for total dose monitoring and with graphite and argon for gamma dose monitoring. This study, in addition, tested the Monte Carlo modeling of the FNIS system, as well as the performance of the system by comparing the calculated results with experimental measurements using activation foils and paired ion chambers.

  3. A target station for plasma exposure of neutron irradiated fusion material samples to reactor relevant conditions

    NASA Astrophysics Data System (ADS)

    Rapp, Juergen; Giuliano, Dominic; Ellis, Ronald; Howard, Richard; Lore, Jeremy; Lumsdaine, Arnold; Lessard, Timothy; McGinnis, William; Meitner, Steven; Owen, Larry; Varma, Venugopal

    2015-11-01

    The Material Plasma Exposure eXperiment (MPEX) is a device planned to address scientific and technological gaps for the development of viable plasma facing components for fusion reactor conditions (FNSF, DEMO). It will have to address the relevant plasma conditions in a reactor divertor (electron density, electron temperature, ion fluxes) and it needs to be able to expose a-priori neutron irradiated samples. A pre design of a target station able to handle activated materials will be presented. This includes detailed MCNP as well as SCALE and MAVRIC calculations for all potential plasma-facing materials to estimate dose rates. Details on the remote handling schemes for the material samples will be presented. 2 point modeling of the linear plasma transport has been used to scope out the parameter range of the anticipated power fluxes to the target. This has been used to design the cooling capability of the target. The operational conditions of surface temperatures, plasma conditions, and oblique angle of incidence of magnetic field to target surface will be discussed. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

  4. 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.

  5. Thermal, structural and mechanical properties of neutron irradiated Bayfol nuclear track detector

    NASA Astrophysics Data System (ADS)

    Nouh, S. A.; Mohamed, Amal; Bahammam, S.

    2009-07-01

    Samples from sheets of the polymeric material Bayfol have been exposed to neutrons of incident energy in the range 0.8-19.2 MeV. The resultant effect of neutron irradiation on the thermal properties of Bayfol has been investigated using thermo-gravimetric analysis. The onset temperature of decomposition and activation energy of thermal decomposition were calculated. The variation of transition temperatures with neutron energy has been determined using differential thermal analysis. The results indicate Bayfol thermograms characterized by the appearance of an endothermic peak due to melting. Melting temperature was found to be dependent on the neutron energy. Structural property studies using infrared spectroscopy were performed and results indicated that scission takes place at the carbonate site with the formation of a hydroxyl group. Mechanical properties were studied and it is shown that, at the fluence range 0-4.4 MeV, the standard chains and a great number of chain ends weaken and the material may become softer.

  6. Genome Resilience and Prevalence of Segmental Duplications Following Fast Neutron Irradiation of Soybean

    PubMed Central

    Bolon, Yung-Tsi; Stec, Adrian O.; Michno, Jean-Michel; Roessler, Jeffrey; Bhaskar, Pudota B.; Ries, Landon; Dobbels, Austin A.; Campbell, Benjamin W.; Young, Nathan P.; Anderson, Justin E.; Grant, David M.; Orf, James H.; Naeve, Seth L.; Muehlbauer, Gary J.; Vance, Carroll P.; Stupar, Robert M.

    2014-01-01

    Fast neutron radiation has been used as a mutagen to develop extensive mutant collections. However, the genome-wide structural consequences of fast neutron radiation are not well understood. Here, we examine the genome-wide structural variants observed among 264 soybean [Glycine max (L.) Merrill] plants sampled from a large fast neutron-mutagenized population. While deletion rates were similar to previous reports, surprisingly high rates of segmental duplication were also found throughout the genome. Duplication coverage extended across entire chromosomes and often prevailed at chromosome ends. High-throughput resequencing analysis of selected mutants resolved specific chromosomal events, including the rearrangement junctions for a large deletion, a tandem duplication, and a translocation. Genetic mapping associated a large deletion on chromosome 10 with a quantitative change in seed composition for one mutant. A tandem duplication event, located on chromosome 17 in a second mutant, was found to cosegregate with a short petiole mutant phenotype, and thus may serve as an example of a morphological change attributable to a DNA copy number gain. Overall, this study provides insight into the resilience of the soybean genome, the patterns of structural variation resulting from fast neutron mutagenesis, and the utility of fast neutron-irradiated mutants as a source of novel genetic losses and gains. PMID:25213171

  7. Neutron irradiation effects in magnesium-aluminate spinel doped with transition metals

    NASA Astrophysics Data System (ADS)

    Gritsyna, V. T.; Afanasyev-Charkin, I. V.; Kobyakov, V. A.; Sickafus, K. E.

    2000-12-01

    We present data on optical properties for stoichiometric (MgO · Al 2O 3) and non-stoichiometric (MgO · 2Al 2O 3) spinel crystals: (1) nominally pure; (2) doped with transition metals Mn, Cr, and Fe to a concentration of 0.01 wt%; (3) irradiated with neutrons to a fluence of 1.8×10 21 m -2; (4) post-annealed at 650 K. The temperature during neutron irradiation was 350 K. Optical absorption and thermoluminescence measurements were performed on irradiated and annealed samples at room temperature. Results of absorption measurements show spectra with the following features: (1) a prominent band at 2.33 eV (for stoichiometric spinel); (2) overlapping bands attributed to hole centers (3.17 eV); (3) optical centers on antisite defects (3.78 and 4.14 eV); (4) F +- and F-centers (4.75 and 5.3 eV); (5) bands related to defect complexes. For nominally pure samples, the efficiency of optical center formation in stoichiometric spinel is half that in non-stoichiometric spinel. Doped crystals exhibit high efficiencies for defect creation, independent of spinel composition. All dopants enhance the efficiency of defect creation in spinel. Doping with Mn has the least effect on increasing the number of radiation-induced stable defects. Apparently, impurities in spinel serve as centers for stabilization of irradiation-induced interstitials or vacancies.

  8. Neutron flux assessment of a neutron irradiation facility based on inertial electrostatic confinement fusion.

    PubMed

    Sztejnberg Gonçalves-Carralves, M L; Miller, M E

    2015-12-01

    Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutron flux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutron flux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutron fluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutron fluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice. PMID:26122974

  9. Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-03-01

    The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.

  10. Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-02-01

    The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.

  11. Neutron-irradiation creep of silicon carbide materials beyond the initial transient

    NASA Astrophysics Data System (ADS)

    Koyanagi, Takaaki; Katoh, Yutai; Ozawa, Kazumi; Shimoda, Kazuya; Hinoki, Tatsuya; Snead, Lance L.

    2016-09-01

    Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. The materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber-reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ∼1 GPa for the fibers and ∼300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ∼1 × 10-7 [dpa-1 MPa-1] at 430-750 °C for the range of 1-30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures-such as grain boundary, crystal orientation, and secondary phases-increase with increasing irradiation temperature.

  12. Segmented Ge detector rejection of internal beta activity produced by neutron irradiation

    NASA Technical Reports Server (NTRS)

    Varnell, L. S.; Callas, J. L.; Mahoney, W. A.; Pehl, R. H.; Landis, D. A.

    1991-01-01

    Future Ge spectrometers flown in space to observe cosmic gamma-ray sources will incorporate segmented detectors to reduce the background from radioactivity produced by energetic particle reactions. To demonstrate the effectiveness of a segmented Ge detector in rejecting background events due to the beta decay of internal radioactivity, a laboratory experiment has been carried out in which radioactivity was produced in the detector by neutron irradiation. A Cf-252 source of neutrons was used to produce, by neutron capture on Ge-74 (36.5 percent of natural Ge) in the detector itself, Ge-75 (t sub 1/2 = 82.78 min), which decays by beta emission with a maximum electron kinetic energy of 1188 keV. By requiring that an ionizing event deposit energy in two or more of the five segments of the detector, each about 1-cm thick, the beta particles, which have a range of about 1-mm, are rejected, while most external gamma rays incident on the detector are counted. Analysis of this experiment indicates that over 85 percent of the beta events from the decay of Ge-75 are rejected, which is in good agreement with Monte Carlo calculations.

  13. 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.

  14. Neutron-irradiation creep of silicon carbide materials beyond the initial transient

    DOE PAGESBeta

    Katoh, Yutai; Ozawa, Kazumi; Shimoda, Kazuya; Hinoki, Tatsuya; Snead, Lance Lewis; Koyanagi, Takaaki

    2016-06-04

    Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. Here, the materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ~1 GPa for the fibers and ~300 MPa for themore » other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ~1 × 10–7 [dpa–1 MPa–1] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.« less

  15. Study of the response of PICASSO bubble detectors to neutron irradiation

    NASA Astrophysics Data System (ADS)

    Marlisov, Daniiar

    The objective of this work was to simulate the PICASSO experiment and to study the detector response to neutron irradiation. The results of the simulation show the rock neutron rate to be 1-2 neutrons/day for the setup used until 2009 and less than 0.1 neutrons/day for the setup used after 2010. The shielding efficiency was calculated to be 98% and 99.6% for the two setups respectively. The detector response to an AmBe source was simulated. Neutron rates differ for two AmBe source spectra from the literature. The observed data rate is in agreement with the rate from the simulation. The detector stability was examined and found to be stable. The source position and orientation affect the detector efficiency creating a systematic uncertainity on the order of 10-35%. This uncertainity was eliminated with a source holder. The localisation of recorded events inside the detector and the simulated neutron distribution agree.

  16. Characterization of neutron calibration fields at the TINT's 50 Ci americium-241/beryllium neutron irradiator

    NASA Astrophysics Data System (ADS)

    Liamsuwan, T.; Channuie, J.; Ratanatongchai, W.

    2015-05-01

    Reliable measurement of neutron radiation is important for monitoring and protection in workplace where neutrons are present. Although Thailand has been familiar with applications of neutron sources and neutron beams for many decades, there is no calibration facility dedicated to neutron measuring devices available in the country. Recently, Thailand Institute of Nuclear Technology (TINT) has set up a multi-purpose irradiation facility equipped with a 50 Ci americium-241/beryllium neutron irradiator. The facility is planned to be used for research, nuclear analytical techniques and, among other applications, calibration of neutron measuring devices. In this work, the neutron calibration fields were investigated in terms of neutron energy spectra and dose equivalent rates using Monte Carlo simulations, an in-house developed neutron spectrometer and commercial survey meters. The characterized neutron fields can generate neutron dose equivalent rates ranging from 156 μSv/h to 3.5 mSv/h with nearly 100% of dose contributed by neutrons of energies larger than 0.01 MeV. The gamma contamination was less than 4.2-7.5% depending on the irradiation configuration. It is possible to use the described neutron fields for calibration test and routine quality assurance of neutron dose rate meters and passive dosemeters commonly used in radiation protection dosimetry.

  17. Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

    NASA Astrophysics Data System (ADS)

    Taylor, C. N.; Shimada, M.; Merrill, B. J.; Akers, D. W.; Hatano, Y.

    2015-08-01

    The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

  18. On He bubbles in neutron irradiated SYLRAMIC type SiC fibers

    SciTech Connect

    Gelles, David S; Youngblood, Gerald E

    2006-03-01

    SylramicTM type SiC fibers, which contain at least 2.3 wt% B, were examined by TEM following neutron irradiation to dose levels of ~7 dpa in HFIR at 800°C and to ~1 dpa in ATR at 1090°C. At these radiation damage dose levels, transmutation of the boron-10 component effectively “dopes” the Sylramic type fibers with up to 10,000 appm helium. Following irradiation at 800°C, bubble development was too fine to resolve even by high resolution TEM. However, following irradiation at 1090°C helium bubble development was resolvable, but complex. A fine dispersion of 1-nm bubbles was observed within the SiC grains and a coarse, non-uniform distribution of irregular 25-nm bubbles was observed on grain boundaries. In addition, some unusual arrays of planar 2.5-nm thick bubbles were observed in the SiC grains and equiaxed bubbles were observed in the boride precipitate particles contained within the fiber microstructure. Not unexpectedly, helium retention and bubble formation in β-SiC depends on details of the polycrystalline microstructure as well as the irradiation conditions.

  19. 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.

  20. Microstructural characterization of deformation localization at small strains in a neutron-irradiated 304 stainless steel

    NASA Astrophysics Data System (ADS)

    Field, Kevin G.; Gussev, Maxim N.; Busby, Jeremy T.

    2014-09-01

    A specific phenomenon - highly localized regions of deformation - was found and investigated at the free surface and near-surface layer of a neutron irradiated AISI 304 stainless steel bend specimen deformed to a maximum surface strain of 0.8%. It was shown that local plastic deformation near the surface might reach significant levels being localized at specific spots even when the maximum free surface strain remains below 1%. The effect was not observed in non-irradiated steel of the same composition at similar strain levels. Cross-sectional EBSD analysis demonstrated that the local misorientation level was highest near the free surface and diminished with increasing depth in these regions. (S)TEM indicated that the local density of dislocation channels might vary up to an order of magnitude. These channels may contain twins or may be twin free depending on grain orientation and local strain levels. BCC-phase (α-martensite) formation associated with channel-grain boundary intersection points was observed using EBSD and STEM in the near-surface layer.

  1. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-05-01

    The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young's modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in "soft" grains with a high Schmid factor located near "stiff" grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to <0 0 1>||TA (tensile axis) and <1 0 1>||TA were twin free and grain with <1 1 1>||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

  2. TEM study of radiation damage and annealing of neutron irradiated zirconolite

    SciTech Connect

    Lumpkin, G.R.; Smith, K.L.; Blake, R.G.

    1996-08-01

    Neutron irradiation was used to simulate alpha-decay damage in zirconolite, resulting in a transformation from the crystalline to the amorphous state at doses of 4--25 {times} 10{sup 19} n/cm{sup 2} (E {ge} 1 MeV). With increasing dose, the radiation damage microstructures resemble damage caused by: (1) alpha-decay of {sup 232}Th and {sup 238}U in natural zirconolites, (2) alpha-decay of {sup 238}Pu or {sup 244}Cm in synthetic samples, and (3) collision cascades in samples irradiated with heavy ions. Heavily damaged zirconolite recovers to a defect fluorite phase on annealing at temperatures up to 1,000 C. The main stage of structural recovery was found to occur at temperatures of 600--800 C. The microstructures after heating depend on the initial level of damage: zirconolite grains with low to moderate levels of damage anneal to imperfect single crystals, whereas heavily damaged grains recrystallize to a polycrystalline microstructure. Complications encountered in this work include the production of fission tracks (due to trace amounts of U) and a non-uniform distribution of damage at higher dose levels (possibly due to electron beam heating).

  3. Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

    SciTech Connect

    Wang, Zujun Huang, Shaoyan; Liu, Minbo; Xiao, Zhigang; He, Baoping; Yao, Zhibin; Sheng, Jiangkun

    2014-07-15

    The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 μm CMOS technology. The flux of neutron beams was about 1.33 × 10{sup 8} n/cm{sup 2}s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 × 10{sup 11}, 5 × 10{sup 11}, and 1 × 10{sup 12} n/cm{sup 2}, respectively. The mean dark signal (K{sub D}), dark signal spike, dark signal non-uniformity (DSNU), noise (V{sub N}), saturation output signal voltage (V{sub S}), and dynamic range (DR) versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike.

  4. 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.

  5. The neutron irradiation effect on the mechanical properties and structure of beryllium

    SciTech Connect

    Fabritsiev, S.A.; Pokrovsky, A.S.; Bagautdinov, R.M.

    1999-10-01

    The neutron irradiation effect on the mechanical properties and structure of beryllium are presented. Irradiation was performed in the BOR-60 reactor up to doses of 0.7--1.1 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) at irradiation temperatures of 350 C, 400 C, 520 C, 780 C. Two modifications of RF beryllium, i.e., DShG-200 and TShG-56, were chosen for investigation. For irradiation at temperatures of 350--400 C Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {le} 300 C. The fracture of samples is of brittle, mainly transcrystallite, type. High-temperature irradiation (T{sub irr} = 780 C) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystalline at T{sub test} {ge} 600 C. Helium embrittlement is also accompanied with a drop in the Be mechanical properties. The conclusion is made that the irradiation temperature range, where irradiated beryllium has a satisfactory level of properties, is rather narrow: 300 C {le} T {le} 500 C.

  6. Kinetics that govern the release of tritium from neutron-irradiated lithium oxide

    SciTech Connect

    Bertone, P.C.

    1986-01-01

    The Lithium Blanket Module (LBM) program being conducted at the Princeton Plasma Physics Laboratory requires that tritium concentrations as low as 0.1 nCi/g, bred in both LBM lithium oxide pellets and gram-size lithium samples, be measured with an uncertainty not exceeding +/-6%. This thesis reports two satisfactory methods of assaying LBM pellets and one satisfactory method of assaying lithium samples. Results of a fundamental kinetic investigation are also reported. The thermally driven release of tritium from neutron-irradiated lithium oxide pellets is studied between the temperatures of 300 and 400/sup 0/C. The observed release clearly obeys first-order kinetics, and the governing activation energy appears to be 28.4 kcal/mole. Finally, a model is presented that may explain the thermally driven release of tritium from a lithium oxide crystal and assemblies thereof. It predicts that under most circumstances the release is controlled by either the diffusion of a tritiated species through the crystal, or by the desorption of tritiated water from it.

  7. Re-weldability of neutron-irradiated stainless steels studied by multi-pass TIG welding

    NASA Astrophysics Data System (ADS)

    Nakata, K.; Oishi, M.; Koshiishi, M.; Hashimoto, T.; Anzai, H.; Saito, Y.; Kono, W.

    2002-12-01

    Weldability of neutron-irradiated stainless steel (SS) has been studied by multi-pass bead-on-plate and build-up tungsten inert gas (TIG) welding, simulating the repair-welding of reactor components. Specimens were submerged arc welding (SAW) joint of Type 304 SS containing 0.5 appm helium (1.8 appm in the SAW weld metal). Sound welding could be obtained by one- to three-pass welding on the plates at weld heat inputs less than 1 MJ/m in the irradiated 304 SS base metal. In the case of the build-up welding of a groove, no visible defects appeared in the specimen at a heat input as low as 0.4 MJ/m. However, build-up welding at a high heat input of 1 MJ/m was prone to weld cracking, owing to the formation of helium bubbles on grain boundaries of the base metal or dendrite boundaries of pre-existing SAW weld metal, in the area within 0.6 mm from the fusion line.

  8. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    DOE PAGESBeta

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For amore » single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.« less

  9. 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.

  10. 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.

  11. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    SciTech Connect

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.

  12. Analytical description of true stress-true strain curves for neutron-irradiated stainless austenitic steels

    SciTech Connect

    Gussev, Maxim N; Byun, Thak Sang; Busby, Jeremy T

    2012-01-01

    This paper summarizes the results of an investigation for the deformation hardening behaviors of neutron-irradiated stainless steels in terms of true stress( ) true strain( ) curves. It is commonly accepted that the - curves are more informative for describing plastic flow, but there are few papers devoted to using the true curves for describing constitutive behaviors of materials. This study uses the true curves obtained from stainless steel samples irradiated to doses in the range of 0 55 dpa by various means: finite element calculation, optic extensomentry, and recalculation of engineering curves. It is shown that for the strain range 0 0.6 the true curves can be well described by the Swift equation: =k ( - 0)0.5. The influence of irradiation on the parameters of the Swift equation is investigated in detail. It is found that in most cases the k-parameter of this equation is not changed significantly by irradiation. Since large data scattering was observed for the 0-parameter, a modified Swift equation =k*( - 0 2/k2)0.5 was proposed and evaluated. This equation is based on the concept of zero stress, which is, in general, close to yield stress. The relationships among k, 0, and damage dose are discussed in detail, so as to more accurately describe the true curves for irradiated stainless steels.

  13. Influence of Heat Treatment on Corrosion Resistance of High-Velocity Oxygen-Fuel Sprayed WC-17Co Coatings on 42CrMo Steel

    NASA Astrophysics Data System (ADS)

    Sun, Wan-chang; Zhang, Pei; Zhang, Feng; Dong, Chu-xu; Zhang, Ju-mei; Cai, Hui

    2015-09-01

    The influence of heat treatment from 500 to 1100 °C on the 5 wt.% H2SO4 solution-induced corrosion resistance of high-velocity oxygen-fuel sprayed WC-17Co coatings on 42CrMo steel was investigated, by using x-ray diffractometer (XRD), scanning electron microscopy (SEM)-energy-dispersive spectrometer (EDS), and polarization curve methods. XRD analysis showed decrease in W2C phase intensity with recrystallization of amorphous Co and generation of new Co3W3C and Co6W6C phases with heat treatment. Porosity distribution did not follow a particular pattern; it initially increased and then decreased with increasing temperature. Corrosion resistance sequence of the as-sprayed and heat-treated coatings in 5 wt.% H2SO4 solution was C-5 > C-9 > C-A > C-7 > C-11. Furthermore, microstructure and phase structure of heat-treated coatings revealed the formation of different discontinuous plate-like oxide films on the surface of the heat-treated coatings which indicated the vital effect of binder structure on the corrosion resistance.

  14. Effects of thermal treatments on microstructure and mechanical properties of a Co-Cr-Mo-W biomedical alloy produced by laser sintering.

    PubMed

    Mengucci, P; Barucca, G; Gatto, A; Bassoli, E; Denti, L; Fiori, F; Girardin, E; Bastianoni, P; Rutkowski, B; Czyrska-Filemonowicz, A

    2016-07-01

    Direct Metal Laser Sintering (DMLS) technology based on a layer by layer production process was used to produce a Co-Cr-Mo-W alloy specifically developed for biomedical applications. The alloy mechanical response and microstructure were investigated in the as-sintered state and after post-production thermal treatments. Roughness and hardness measurements, and tensile and flexural tests were performed to study the mechanical response of the alloy while X-ray diffraction (XRD), electron microscopy (SEM, TEM, STEM) techniques and microanalysis (EDX) were used to investigate the microstructure in different conditions. Results showed an intricate network of ε-Co (hcp) lamellae in the γ-Co (fcc) matrix responsible of the high UTS and hardness values in the as-sintered state. Thermal treatments increase volume fraction of the ε-Co (hcp) martensite but slightly modify the average size of the lamellar structure. Nevertheless, thermal treatments are capable of producing a sensible increase in UTS and hardness and a strong reduction in ductility. These latter effects were mainly attributed to the massive precipitation of an hcp Co3(Mo,W)2Si phase and the contemporary formation of Si-rich inclusions. PMID:26803005

  15. Mechanical and thermal properties of h-MX{sub 2} (M = Cr, Mo, W; X = O, S, Se, Te) monolayers: A comparative study

    SciTech Connect

    Çakır, Deniz Peeters, François M.; Sevik, Cem

    2014-05-19

    Using density functional theory, we obtain the mechanical and thermal properties of MX{sub 2} monolayers (where M = Cr, Mo, W and X = O, S, Se, Te). The Γ-centered phonon frequencies (i.e., A{sub 1}, A{sub 2}{sup ″}, E′, and E″), relative frequency values of A{sub 1}, and E′ modes, and mechanical properties (i.e., elastic constants, Young modulus, and Poisson's ratio) display a strong dependence on the type of metal and chalcogenide atoms. In each chalcogenide (metal) group, transition-metal dichalcogenides (TMDCs) with W (O) atom are found to be much stiffer. Consistent with their stability, the thermal expansion of lattice constants for TMDCs with O (Te) is much slower (faster). Furthermore, in a heterostructure of these materials, the difference of the thermal expansion of lattice constants between the individual components becomes quite tiny over the whole temperature range. The calculated mechanical and thermal properties show that TMDCs are promising materials for heterostructures.

  16. Atom probe study of the carbon distribution in a hardened martensitic hot-work tool steel X38CrMoV5-1.

    PubMed

    Lerchbacher, Christoph; Zinner, Silvia; Leitner, Harald

    2012-07-01

    The microstructure of the hardened common hot-work tool steel X38CrMoV5-1 has been characterized by atom probe tomography with the focus on the carbon distribution. Samples quenched with technically relevant cooling parameters λ from 0.1 (30 K/s) to 12 (0.25 K/s) have been investigated. The parameter λ is an industrially commonly used exponential cooling parameter, representing the cooling time from 800 to 500 °C in seconds divided with hundred. In all samples pronounced carbon segregation to dislocations and cluster formation could be observed after quenching. Carbon enriched interlath films with peak carbon levels of 6-10 at.%, which have been identified to be retained austenite by TEM, show a thickness increase with increasing λ. Therefore, the fraction of total carbon staying in the austenite grows. This carbon is not available for the tempering induced precipitation of secondary carbides in the bulk. Through all samples no segregation of any substitutional elements takes place. Charpy impact testing and fracture surface analysis of the hardened samples reveal the cooling rate induced microstructural distinctions. PMID:22391101

  17. Electrochemical comparison and biological performance of a new CoCrNbMoZr alloy with commercial CoCrMo alloy.

    PubMed

    Andrei, M; Galateanu, B; Hudita, A; Costache, M; Osiceanu, P; Calderon Moreno, J M; Drob, S I; Demetrescu, I

    2016-02-01

    A new CoCrNbMoZr alloy, with Nb and Zr content is characterized from the point of view of surface features, corrosion resistance and biological performance in order to be proposed as dental restorative material. Its properties are discussed in comparison with commercial Heraenium CE alloy based on Co, Cr and Mo as well. The microstructure of both alloys was revealed by scanning electron microscopy (SEM). The composition and thickness of the alloy native passive films were identified by X-ray photoelectron spectroscopy (XPS). The surface characteristics were analyzed by atomic force microscopy (AFM) and contact angle techniques. The quantity of ions released from alloys in artificial saliva was evaluated with inductively coupled plasma-mass spectroscopy (ICP-MS) measurements. The electrochemical stability was studied in artificial Carter-Brugirard saliva, performing open circuit potentials, polarization resistances and corrosion currents and rates. The biological performance of the new alloy was tested in vitro in terms of human adipose stem cells (hASCs) morphology, viability and proliferation status. The new alloy is very resistant to the attack of the aggressive ions from the artificial saliva. The surface properties, the roughness and wettabiliy sustain the cell behavior. The comparison of the new alloy behavior with that of existing commercial CoCrMo alloy showed the superior properties of the new metallic biomaterial. PMID:26652383

  18. The Effect of Temperature, Specimen Size, and Geometry on the Fracture Toughness of a 3 Pct NiCrMoV Low Pressure Turbine Disc Steel

    NASA Astrophysics Data System (ADS)

    Shaw, N. B.; Spink, G. M.

    1983-03-01

    The variation of fracture toughness with temperature and specimen size of a 3 pct NiCrMoV LP disc steel has been investigated over the temperature range -100 to +100 °C using compact tension and single-edge-notched bend geometries. A number of large ‘half-disc’ three point bend specimens were also tested. Toughness increased up to a transition temperature coinciding with the onset of stable ductile tearing prior to instability. Below this temperature fracture could be described by established linear elastic or post yield fracture analyses. Above this temperature failure was by plastic collapse. The transition temperature decreased with decreasing specimen size, and at similar thicknesses was lower for the bend geometry than for the compact tension so that it was not possible to predict the fracture behavior of the full size service component from small scale tests in the transition region. A further complicating feature was the extreme scatter of some duplicate test results below the transition temperature. The implications for toughness testing in the transition region are discussed. The data obtained in this work have been combined with published data for similar steels to derive an equation which describes the variation of fracture toughness with temperature for steels of this type.

  19. Microstructures and mechanical properties of metastable Ti-30Zr-(Cr, Mo) alloys with changeable Young's modulus for spinal fixation applications.

    PubMed

    Zhao, Xiaoli; Niinomi, Mitsuo; Nakai, Masaaki; Miyamoto, Goro; Furuhara, Tadashi

    2011-08-01

    In order to develop a novel alloy with a changeable Young's modulus for spinal fixation applications, we investigated the microstructures, Young's moduli, and tensile properties of metastable Ti-30Zr-(Cr, Mo) alloys subjected to solution treatment (ST) and cold rolling (CR). All the alloys comprise a β phase and small athermal ω phase, and they exhibit low Young's moduli after ST. During CR, deformation-induced phase transformation occurs in all the alloys. The change in Young's modulus after CR is highly dependent on the type of deformation-induced phase. The increase in Young's modulus after CR is attributed to the deformation-induced ω phase on {3 3 2} mechanical twinning. Ti-30Zr-3Cr-3Mo (3Cr3Mo), which exhibits excellent tensile properties and a changeable Young's modulus, shows a smaller springback than Ti-29Nb-13Ta-4.6Zr, a β-type titanium alloy expected to be useful in spinal fixation applications. Thus, 3Cr3Mo is a potential candidate for spinal fixation applications. PMID:21569873

  20. NiAl-based Polyphase in situ Composites in the NiAl-Ta-X (X = Cr, Mo, or V) Systems

    NASA Technical Reports Server (NTRS)

    Johnson, D. R.; Oliver, B. F.; Noebe, R. D.; Whittenberger, J. D.

    1995-01-01

    Polyphase in situ composites were generated by directional solidification of ternary eutectics. This work was performed to discover if a balance of properties could be produced by combining the NiAl-Laves phase and the NiAl-refractory metal phase eutectics. The systems investigated were the Ni-Al-Ta-X (X = Cr, Mo, or V) alloys. Ternary eutectics were found in each of these systems and the eutectic composition, temperature, and morphology were determined. The ternary eutectic systems examined were the NiAl-NiAlTa-(Mo, Ta), NiAl-(Cr, Al) NiTa-Cr, and the NiAl-NiAlTa-V systems. Each eutectic consists of NiAl, a C14 Laves phase, and a refractory metal phase. Directional solidification was performed by containerless processing techniques in a levitation zone refiner to minimize alloy contamination. Room temperature fracture toughness of these materials was determined by a four-point bend test. Preliminary creep behavior was determined by compression tests at elevated temperatures, 1100-l400 K. Of the ternary eutectics, the one in the NiAl-Ta-Cr system was found to be the most promising. The fracture toughness of the NiAl-(Cr, Al)NiTa-Cr eutectic was intermediate between the values of the NiAl-NiAlTa eutectic and the NiAl-Cr eutectic. The creep strength of this ternary eutectic was similar to or greater than that of the NiAl-Cr eutectic.

  1. Theoretical predictions of properties and gas-phase chromatography behaviour of carbonyl complexes of group-6 elements Cr, Mo, W, and element 106, Sg.

    PubMed

    Pershina, V; Anton, J

    2013-05-01

    Fully relativistic, four-component density functional theory electronic structure calculations were performed for M(CO)6 of group-6 elements Cr, Mo, W, and element 106, Sg, with an aim to predict their adsorption behaviour in the gas-phase chromatography experiments. It was shown that seaborgium hexacarbonyl has a longer M-CO bond, smaller ionization potential, and larger polarizability than the other group-6 molecules. This is explained by the increasing relativistic expansion and destabilization of the (n - 1)d AOs with increasing Z in the group. Using results of the calculations, adsorption enthalpies of the group-6 hexacarbonyls on a quartz surface were predicted via a model of physisorption. According to the results, -ΔHads should decrease from Mo to W, while it should be almost equal--within the experimental error bars--for W and Sg. Thus, we expect that in the future gas-phase chromatography experiments it will be almost impossible--what concerns ΔHads--to distinguish between the W and Sg hexacarbonyls by their deposition on quartz. PMID:23656128

  2. Theoretical predictions of properties and gas-phase chromatography behaviour of carbonyl complexes of group-6 elements Cr, Mo, W, and element 106, Sg

    NASA Astrophysics Data System (ADS)

    Pershina, V.; Anton, J.

    2013-05-01

    Fully relativistic, four-component density functional theory electronic structure calculations were performed for M(CO)6 of group-6 elements Cr, Mo, W, and element 106, Sg, with an aim to predict their adsorption behaviour in the gas-phase chromatography experiments. It was shown that seaborgium hexacarbonyl has a longer M-CO bond, smaller ionization potential, and larger polarizability than the other group-6 molecules. This is explained by the increasing relativistic expansion and destabilization of the (n - 1)d AOs with increasing Z in the group. Using results of the calculations, adsorption enthalpies of the group-6 hexacarbonyls on a quartz surface were predicted via a model of physisorption. According to the results, -ΔHads should decrease from Mo to W, while it should be almost equal - within the experimental error bars - for W and Sg. Thus, we expect that in the future gas-phase chromatography experiments it will be almost impossible - what concerns ΔHads - to distinguish between the W and Sg hexacarbonyls by their deposition on quartz.

  3. Effect of synovial fluid, phosphate-buffered saline solution, and water on the dissolution and corrosion properties of CoCrMo alloys as used in orthopedic implants.

    PubMed

    Lewis, A C; Kilburn, M R; Papageorgiou, I; Allen, G C; Case, C P

    2005-06-15

    The corrosion and dissolution of high- and low-carbon CoCrMo alloys, as used in orthopedic joint replacements, were studied by immersing samples in phosphate-buffered saline (PBS), water, and synovial fluid at 37 degrees C for up to 35 days. Bulk properties were analyzed with a fine ion beam microscope. Surface analyses by X-ray photoelectron spectroscopy and Auger electron spectroscopy showed surprisingly that synovial fluid produced a thin oxide/hydroxide layer. Release of ions into solution from the alloy also followed an unexpected pattern where synovial fluid, of all the samples, had the highest Cr concentration but the lowest Co concentration. The presence of carbide inclusions in the alloy did not affect the corrosion or the dissolution mechanisms, although the carbides were a significant feature on the metal surface. Only one mechanism was recognized as controlling the thickness of the oxide/hydroxide interface. The analysis of the dissolved metal showed two mechanisms at work: (1) a protein film caused ligand-induced dissolution, increasing the Cr concentration in synovial fluid, and was explained by the equilibrium constants; (2) corrosion at the interface increased the Co in PBS. The effect of prepassivating the samples (ASTM F-86-01) did not always have the desired effect of reducing dissolution. The release of Cr into PBS increased after prepassivation. The metal-synovial fluid interface did not contain calcium phosphate as a deposit, typically found where samples are exposed to calcium rich bodily fluids. PMID:15900610

  4. Corrosion and Fretting Corrosion Studies of Medical Grade CoCrMo Alloy in a Clinically Relevant Simulated Body Fluid Environment

    NASA Astrophysics Data System (ADS)

    Ocran, Emmanuel K.; Guenther, Leah E.; Brandt, Jan-M.; Wyss, Urs; Ojo, Olanrewaju A.

    2015-06-01

    In modular hip implants, fretting corrosion at the head/neck and neck/stem interfaces has been identified as a major cause of early revision in hip implants, particularly those with heads larger than 32 mm. It has been found that the type of fluid used to simulate the fretting corrosion of biomedical materials is crucial for the reliability of laboratory tests. Therefore, to properly understand and effectively design against fretting corrosion damage in modular hips, there is the need to replicate the human body environment as closely as possible during in vitro testing. In this work, corrosion and fretting corrosion behavior of CoCrMo in 0.14 M NaCl, phosphate buffered saline, and in a clinically relevant novel simulated body fluid was studied using a variety of electrochemical characterization techniques and tribological experiments. Electrochemical, spectroscopy and tribo-electrochemical techniques employed include Potentiodynamic polarization, Potentiostatic polarization, Electrochemical impedance spectroscopy, X-ray photoelectron spectroscopy, augur electron spectroscopy, inductively coupled plasma mass spectroscopy, and pin-on-disk wear simulation. The presence of phosphate ions in PBS accounted for the higher corrosion rate when compared with 0.14 M NaCl and the clinically relevant novel simulated body fluid. The low corrosion rates and the nature of the protective passive film formed in the clinically relevant simulated body fluid make it suitable for future corrosion and fretting corrosion studies.

  5. High-resolution diffraction for residual stress determination in the NiCrMoV wheel of an axial compressor for a heavy-duty gas turbine

    NASA Astrophysics Data System (ADS)

    Rogante, M.; Török, G.; Ceschini, G. F.; Tognarelli, L.; Füzesy, I.; Rosta, L.

    2004-07-01

    The wheel of an axial compressor for a heavy-duty gas turbine has been investigated for residual stresses (RS) evaluation of the teeth-section where SANS measurements have previously been performed. Such a component can contain internal RS, either due to the manufacturing process, or to the operating cycles fatigue. The constitutive material is a NiCrMoV steel to ASTM A 471 (type 2) norms (equivalent to B50A420B10); this material is usually adopted in the manufacturing of forged components for gas turbines. Internal radial and hoop RS have been determined, whose values are under the limit of 200kPa. Hoop RS, in general, resulted in higher value than the radial ones. The present experiment represents a particularly important step in the RS determination for gas turbine components, since the measurements reveal that the fatigue of the wheel is also a lifetime limiting factor although, in the same technological field, the available data in the actual neutron techniques literature mainly concern turbine buckets.

  6. Nanoscale phase separation in epitaxial Cr-Mo and Cr-V alloy thin films studied using atom probe tomography: Comparison of experiments and simulation

    SciTech Connect

    Devaraj, A.; Ramanan, S.; Walvekar, S.; Bowden, M. E.; Shutthanandan, V.; Kaspar, T. C.; Kurtz, R. J.

    2014-11-21

    Tailored metal alloy thin film-oxide interfaces generated using molecular beam epitaxy (MBE) deposition of alloy thin films on a single crystalline oxide substrate can be used for detailed studies of irradiation damage response on the interface structure. However, the presence of nanoscale phase separation in the MBE grown alloy thin films can impact the metal-oxide interface structure. Due to nanoscale domain size of such phase separation, it is very challenging to characterize by conventional techniques. Therefore, laser assisted atom probe tomography (APT) was utilized to study the phase separation in epitaxial Cr{sub 0.61}Mo{sub 0.39}, Cr{sub 0.77}Mo{sub 0.23}, and Cr{sub 0.32}V{sub 0.68} alloy thin films grown by MBE on MgO(001) single crystal substrates. Statistical analysis, namely frequency distribution analysis and Pearson coefficient analysis of experimental data was compared with similar analyses conducted on simulated APT datasets with known extent of phase separation. Thus, the presence of phase separation in Cr-Mo films, even when phase separation was not clearly observed by x-ray diffraction, and the absence of phase separation in the Cr-V film were confirmed.

  7. Nanoscale Phase Separation In Epitaxial Cr-Mo and Cr-V Alloy Thin Films Studied Using Atom Probe Tomography. Comparison Of Experiments And Simulation

    SciTech Connect

    Devaraj, Arun; Kaspar, Tiffany C.; Ramanan, Sathvik; Walvekar, Sarita K.; Bowden, Mark E.; Shutthanandan, V.; Kurtz, Richard J.

    2014-11-21

    Tailored metal alloy thin film-oxide interfaces generated using molecular beam epitaxial (MBE) deposition of alloy thin films on a single crystalline oxide substrate can be used for detailed studies of irradiation damage response on the interface structure. However presence of nanoscale phase separation in the MBE grown alloy thin films can impact the metal-oxide interface structure. Due to nanoscale domain size of such phase separation it is very challenging to characterize by conventional techniques. Therefor laser assisted atom probe tomography (APT) was utilized to study the phase separation in epitaxial Cr0.61Mo0.39, Cr0.77Mo0.23, and Cr0.32V0.68 alloy thin films grown by MBE on MgO(001) single crystal substrates. Statistical analysis, namely frequency distribution analysis and Pearson coefficient analysis of experimental data was compared with similar analyses conducted on simulated APT datasets with known extent of phase separation. Thus the presence of phase separation in Cr-Mo films, even when phase separation was not clearly observed by x-ray diffraction, and the absence of phase separation in the Cr-V film were thus confirmed.

  8. 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.

  9. Putting chromium on the map for N2 reduction: production of hydrazine and ammonia. A study of cis-M(N2)2 (M = Cr, Mo, W) bis(diphosphine) complexes.

    PubMed

    Egbert, Jonathan D; O'Hagan, Molly; Wiedner, Eric S; Bullock, R Morris; Piro, Nicholas A; Kassel, W Scott; Mock, Michael T

    2016-07-19

    The first complete structurally and spectroscopically characterized series of isostructural Group 6 N2 complexes is reported. Protonolysis experiments on cis-[M(N2)2(P(Et)N(R)P(Et))2] (M = Cr, Mo, W; R = 2,6-difluorobenzyl) reveal that only Cr affords N2H5(+) and NH4(+) from the reduction of the N2 ligands. PMID:27331373

  10. Transformation, metallurgical response and behavior of the weld fusion zone and heat affected zone in Cr-Mo steels for fossil energy application: Final technical report for January 1985-September 1987

    SciTech Connect

    Lundin, C.D.; Henning, J.A.; Menon, R.; Khan, K.K.

    1987-09-30

    This research program was undertaken to provide fundamental and basic metallurgical information on the behavior of the heat affected zone (HAZ) in Cr-Mo steel welds as well as practical information on their relative weldability. The principal work was the evaluation of the post weld heat treatment (PWHT) cracking of Cr-Mo steels ranging in Cr content from 2-1/4% to 9%. Differences in observed cracking behavior were contrasted with composition, on cooling transformation behavior and HAZ microstructure. Hydrogen assisted cracking (HAC) studies using a large scale cracking test were conducted on 2-1/4 Cr and 3 Cr steels. Soft zone studies were conducted on 9 Cr NKK steel to determine the reason for the development of a low hardness region (''Soft Zone'') at the outer boundary of the HAZ. The literature review provides a concise historical review and the basis of theories for PWHT cracking and HAC in Cr-Mo steels which were employed to explain the weld cracking susceptibility of various Cr-Mo alloys. PWHT cracking susceptibility was investigated using Gleeble simulated heat affected zone (HAZ) specimens. A new test was developed at the University of Tennessee, The C-ring test, to evaluate the PWHT cracking behavior. The C-ring test was found to be an extremely useful test for PWHT cracking susceptibility and for verifying the results obtained from Gleeble tsting. An excellent correlation was obtained for the two tests. The standard Y-groove test was selected for HAC susceptibility testing. This test is very suitable for evaluating the HAC of the base metal and is defined in Japanese Industrial Standard JIS Z 3158.

  11. Stress Corrosion Cracking of Ferritic Materials for Fossil Power Generation Applications

    SciTech Connect

    Pawel, Steven J; Siefert, John A.

    2014-01-01

    Creep strength enhanced ferritic (CSEF) steels Grades 23, 24, 91, and 92 have been widely implemented in the fossil fired industry for over two decades. The stress corrosion cracking (SCC) behavior of these materials with respect to mainstay Cr-Mo steels (such as Grades 11, 12 and 22) has not been properly assessed, particularly in consideration of recent reported issues of SCC in CSEF steels. This report details the results of Jones test exposures of a wide range of materials (Grades 11, 22, 23, 24, and 92), material conditions (as-received, improper heat treatments, normalized, weldments) and environments (salt fog; tube cleaning environments including decreasing, scale removal, and passivation; and high temperature water) to compare the susceptibility to cracking of these steels. In the as-received (normalized and tempered) condition, none of these materials are susceptible to SCC in the environments examined. However, in the hardened condition, certain combinations of environment and alloy reveal substantial SCC susceptibility.

  12. Effect of aging temperature on the microstructures and mechanical properties of ZG12Cr9Mo1Co1NiVNbNB ferritic heat-resistant steel

    NASA Astrophysics Data System (ADS)

    Yang, Xue; Sun, Lan; Xiong, Ji; Zhou, Ping; Fan, Hong-yuan; Liu, Jian-yong

    2016-02-01

    The effect of aging on the mechanical properties and microstructures of a new ZG12Cr9Mo1Co1NiVNbNB ferritic heat resistant steel was investigated in this work to satisfy the high steam parameters of the ultra-supercritical power plant. The results show that the main precipitates during aging are Fe(Cr, Mo)23C6, V(Nb)C, and (Fe2Mo) Laves in the steel. The amounts of the precipitated phases increase during aging, and correspondingly, the morphologies of phases are similar to be round. Fe(Cr, Mo)23C6 appears along boundaries and grows with increasing temperature. In addition, it is revealed that the martensitic laths are coarsened and eventually happen to be polygonization. The hardness and strength decrease gradually, whereas the plasticity of the steel increases. What's more, the hardness of this steel after creep is similar to that of other 9%-12%Cr ferritic steels. Thus, ZG12Cr9Mo1Co1NiVNbNB can be used in the project.

  13. Dimensional isotropy of 6H and 3C SiC under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Snead, Lance L.; Katoh, Yutai; Koyanagi, Takaaki; Terrani, Kurt; Specht, Eliot D.

    2016-04-01

    This investigation experimentally determines the as-irradiated crystal axes dimensional change of the common polytypes of SiC considered for nuclear application. Single crystal α-SiC (6H), β-SiC (3C), CVD β-SiC, and single crystal Si have been neutron irradiated near 60 °C from 2 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.02-20 dpa, in order to study the effect of irradiation on bulk swelling and strain along independent crystalline axes. Single crystal, powder diffractometry and density measurement have been carried out. For all neutron doses where the samples remained crystalline all SiC materials demonstrated equivalent swelling behavior. Moreover the 6H-SiC expanded isotropically. The magnitude of the swelling followed a ∼0.77 power law against dose consistent with a microstructure evolution driven by single interstitial (carbon) mobility. Extraordinarily large ∼7.8% volume expansion in SiC was observed prior to amorphization. Above ∼0.9 × 1025 n/m2 (E > 0.1 MeV) all SiC materials became amorphous with an identical swelling: a 11.7% volume expansion, lowering the density to 2.84 g/cm3. The as-amorphized density was the same at the 2 × 1025 and 2 × 1026 n/m2 (E > 0.1 MeV) dose levels.

  14. Mechanical- and physical-property changes of neutron-irradiated chemical-vapor-deposited silicon carbide

    SciTech Connect

    Osborne, M.C.; Steiner, D.; Hay, J.C.; Snead, L.L.

    1999-09-01

    Indentation and density measurements have revealed important changes in the mechanical and physical properties of silicon carbide (SiC) due to neutron irradiation. Specifically, the changes in the elastic modulus, hardness, fracture toughness, and density with irradiation have provided an understanding of the expected performance of SiC and SiC composites in nuclear applications. After the accumulated damage has saturated, these mechanical properties were affected primarily by the irradiation temperature. Chemical-vapor-deposited (CVD) SiC was irradiated above the saturation fluence and yielded volumetric swelling of 2.6% and 1.3% for irradiation temperatures of 100--150 C and 500--550 C, respectively. At the same respective temperatures, the elastic modulus decreased from an unirradiated value of 503 GPa to {approximately} 420 and 450 GPa. Conversely, the hardness increased from 36 GPa for the unirradiated at 100--150 C and 500--550 C, respectively. Interestingly, these two independent properties approached almost-constant levels after exposure to a fluence of 0.5 {times} 10{sup 25} n/m{sup 2}, E > 0.1 MeV. Indentation fracture toughness measurements, which were within the range of values in the literature for conventional fracture toughness procedures for SiC, increased from {approximately} 2.8 MPa{center_dot}m{sup 1/2} for the unirradiated samples to 3.7 and 4.2 MPa{center_dot}m{sup 1/2} for the samples that were irradiated at 100--150 C and 500--550 C, respectively.

  15. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    DOE PAGESBeta

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Finalmore » thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².« less

  16. 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.

  17. Dimensional isotropy of 6H and 3C SiC under neutron irradiation

    DOE PAGESBeta

    Snead, Lance L.; Katoh, Yutai; Koyanagi, Takaaki; Terrani, Kurt A.; Specht, Eliot D.

    2016-01-16

    This investigation experimentally determines the as-irradiated crystal axes dimensional change of the common polytypes of SiC considered for nuclear application. Single crystal α-SiC (6H), β-SiC (3C), CVD β-SiC, and single crystal Si have been neutron irradiated near 60 °C from 2 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.02–20 dpa, in order to study the effect of irradiation on bulk swelling and strain along independent crystalline axes. Single crystal, powder diffractometry and density measurement have been carried out. For all neutron doses where the samples remained crystalline all SiC materials demonstrated equivalent swelling behavior.more » Moreover the 6H–SiC expanded isotropically. The magnitude of the swelling followed a ~0.77 power law against dose consistent with a microstructure evolution driven by single interstitial (carbon) mobility. Extraordinarily large ~7.8% volume expansion in SiC was observed prior to amorphization. Above ~0.9 × 1025 n/m2 (E > 0.1 MeV) all SiC materials became amorphous with an identical swelling: a 11.7% volume expansion, lowering the density to 2.84 g/cm3. As a result, the as-amorphized density was the same at the 2 × 1025 and 2 × 1026 n/m2 (E > 0.1 MeV) dose levels.« less

  18. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    DOE PAGESBeta

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Wirth, Brian D; Snead, Lance Lewis

    2016-01-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutronmore » irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.« less

  19. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    SciTech Connect

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².

  20. Study of the VMM1 read-out chip in a neutron irradiation environment

    NASA Astrophysics Data System (ADS)

    Alexopoulos, T.; Fanourakis, G.; Geralis, T.; Kokkoris, M.; Kourkoumeli-Charalampidi, A.; Papageorgiou, K.; Tsipolitis, G.

    2016-05-01

    Within 2015, the LHC operated close to the design energy of √s = 13–14 TeV delivering instantaneous luminosities up to Script L = 5 × 1033 cm‑2s‑1. The ATLAS Phase-I upgrade in 2018/19 will introduce the MicroMEGAS detectors in the area of the small wheel at the end caps. Accompanying new electronics are designed and built such as the VMM front end ASIC, which provides energy, timing and triggering information and allows fast data read-out. The first VMM version (VMM1) has been widely produced and tested in various test beams, whilst the second version (VMM2) is currently being tested. This paper focuses on the VMM1 single event upset studies and more specifically on the response of the configuration registers under harsh radiation environments. Similar conditions are expected at Run III with Script L = 2 × 1034 cm‑2s‑1 and a mean of 55 interactions per bunch crossing. Two VMM1s were exposed in a neutron irradiation environment using the TANDEM Van Der Graaff accelerator at NSCR Demokritos, Athens, Greece. The results showed a rate of SEU occurrences at a measured cross section of (4.1±0.8)×10‑14 cm2/bit for each VMM. Consequently, when extrapolating this value to the luminosity expected in Run III, the occurrence is roughly 6 SEUs/min in all the read-out system comprising 40,000 VMMs installed during the Phase-I upgrade.

  1. Metal-Centered 17-Electron Radicals CpM(CO)3• (M = Cr, Mo, W): A Combined Negative Ion Photoelectron Spectroscopic and Theoretical Study

    SciTech Connect

    van der Eide, Edwin F.; Hou, Gao-Lei; Deng, Shihu; Wen, Hui; Yang, Ping; Bullock, R. Morris; Wang, Xue B.

    2013-04-08

    Despite the importance of group VI metal-centered 17-electron radicals CpM(CO)3 (Cp = 5-C5H5, M = Cr, Mo, W) in establishing many of the fundamental reactions now known for metal-centered radicals, spectroscopic characterization of their electronic properties and structures has been very challenging due to their high reactivity. Here we report a gas-phase study of these species by means of photodetachment photoelectron spectroscopy (PES) of their corresponding 18-electron anions and theoretical electronic structure calculations. Three well-separated spectral features are observed by PES for each anionic species. Electron affinities (EAs) of CpM(CO)3 were experimentally measured from the threshold of each spectrum to be 2.38 ± 0.02 (M = Cr), 2.63 ± 0.02 (Mo), and 2.63 ± 0.01 eV for (W), well correlated with the reported redox potentials measured in solution. Theoretical calculations for all anionic and neutral (radical) species gave calculated EAs and band gaps that are in good agreement with the experimental data. Molecular orbital (MO) analyses for each anion indicate that the top three occupied MOs are mainly metal-based and contribute to the first spectral feature, whereas the next two MOs are largely from C5H5 moiety and contribute to the second spectral feature. The calculations further exhibit appreciable anion-to-neutral structural changes for all three species but with the change for the W species being the smallest, consistent with the W spectra being better resolved than the other two.

  2. High-Temperature Oxidation of Cr-Mo Steels and Its Relevance to Accelerated Rupture Testing and Life Assessment of In-Service Components

    NASA Astrophysics Data System (ADS)

    Singh Raman, R. K.; Al-Mazrouee, A.

    2007-08-01

    Use of accelerated creep rupture testing to assess the remaining life of components operating at elevated temperatures, such as pipes and tubes, is a common practice. At high temperatures, oxide growth can affect the creep results by diameter reduction and thus can increase the stress. However, the nature of oxide layer and hence oxidation behavior can be affected by minor changes in alloying composition of steels. This article presents the study of oxide-scale growth and specimen diameter reduction kinetics during oxidation of two Cr-Mo steels used in the manufacture of boiler tubing. Oxidation tests were carried out on 1.25Cr-0.5Mo and 2.25Cr-1Mo steels at 600 °C and 700 °C for times up to 1000 hours, using cylindrical specimens (similar to those used for creep testing). At 600 °C, the oxidation resistance of 2.25Cr-1Mo steel was superior to 1.25Cr-0.5Mo steel. However, the oxidation resistance of the two steels at 700 °C was similar in spite of the difference in their Cr contents. Multilayer oxide scales of oxides with various compositions were observed to have formed over the two steels. The similarity in oxidation kinetics of the two steels at 700 °C (in spite of differences in Cr contents) is ascribed to their Si contents and the predominant role of Si in oxidation at this temperature. The article also discusses implications of the variation in the oxidation kinetics to the stress enhancement in creep specimens due to scaling losses, and possible inaccuracies in creep data, as a result of minor variations in alloying composition.

  3. Caustic stress corrosion cracking of NiCrMoV rotor steels—The effects of impurity segregation and variation in alloy composition

    NASA Astrophysics Data System (ADS)

    Bandyopadhyay, N.; Briant, C. L.

    1983-10-01

    This paper reports a study of the effects of phosphorus, tin, and molybdenum on the caustic stress corrosion cracking susceptibility of NiCrMoV rotor steels. Constant load tests were performed on these steels in 9M NaOH at 98 ± 1 °C at a controlled potential of either -800 mVHg/Hgo or -400 mVHg/Hgo. Times to failure were measured. The results show that at a potential of -400 mVHg/Hgo the segregation of phosphorus to grain boundaries lowers the resistance of these steels to caustic stress corrosion cracking. When molybdenum is removed from a steel that has phosphorus segregated to the grain boundaries, the steel’s resistance to stress corrosion cracking is improved. High purity alloys, both with and without molybdenum, show very good resistance to caustic cracking at this potential. At-800 mVHg/Hgo segregated phophorus has no effect; only molybdenum additions lower the resistance of the steel to caustic stress corrosion cracking. Segregated tin has little effect at either potential. Metallographic examination shows that one explanation for these results is that molybdenum and phosphorus, probably as anions precipitated from solution, aid in passivating the sides of the crack and thus help keep the crack tip sharp. This sharpness will increase the speed with which the crack will propagate through the sample. Furthermore, removal of molybdenum greatly increases the number of cracks which nucleate. This higher crack density would increase the relative area of the anode to the cathode and thus act to decrease the crack growth rate.

  4. Cold-rolling behavior of biomedical Ni-free Co-Cr-Mo alloys: Role of strain-induced ε martensite and its intersecting phenomena.

    PubMed

    Mori, Manami; Yamanaka, Kenta; Chiba, Akihiko

    2015-03-01

    Ni-free Co-Cr-Mo alloys are some of the most difficult-to-work metallic materials used commonly in biomedical applications. Since the difficulty in plastically deforming them limits their use, an in-depth understanding of their plastic deformability is of crucial importance for both academic and practical purposes. In this study, the microstructural evolution of a Co-29Cr-6Mo-0.2N (mass%) alloy during cold rolling was investigated. Further, its work-hardening behavior is discussed while focusing on the strain-induced face-centered cubic (fcc) γ→hexagonal close-packed (hcp) ε martensitic transformation (SIMT). The planar dislocation slip and subsequent SIMT occurred even in the initial stage of the deformation process owing to the low stability of the γ-phase and contributed to the work hardening behavior. However, the amount of the SIMTed ε-phase did not explain the overall variation in work hardening during cold rolling. It was found that the intersecting of the SIMTed ε-plates enhanced local strain evolution and then produced fine domain-like deformation microstructures at the intersections. Consequently, the degree of work hardening was reduced during subsequent plastic deformation, resulting in the alloy exhibiting a two-stage work hardening behavior. The results obtained in this study suggest that the interaction between ε-martensites, and ultimately its relaxation mechanism, is of significant importance; therefore, this aspect should be addressed in detail; the atomic structures of the γ-matrix/ε-martensite interfaces, the phenomenon of slip transfer at the interfaces, and the slipping behavior of the ε-phase itself are needed to be elucidated for further increasing the cold deformability of such alloys. PMID:26594780

  5. Transition Metal Complexes of Cr, Mo, W and Mn Containing {eta}{sup 1}(S)-2,5-Dimethylthiophene, Benzothiophene and Dibenzothiophene Ligands

    SciTech Connect

    Reynolds, M.

    2000-09-21

    The UV photolysis of hexanes solutions containing the complexes M(CO){sub 6} (M=Cr, Mo, W) or CpMn(CO){sub 3} (Cp={eta}{sup 5}-C{sub 5}H{sub 5}) and excess thiophene (T{sup *}) (T{sup *}=2,5-dimethylthiophene (2,5-Me{sub 2}T), benzothiophene (BT), and dibenzothiophene (DBT)) produces the {eta}{sup 1}(S)-T{sup *} complexes (CO){sub 5}M({eta}{sup 1}(S)-T{sup *}) 1-8 or Cp(CO){sub 2}Mn({eta}{sup 1}(S)-T{sup *})9-11, respectively. However, when T{sup *}=DBT, and M=Mo, a mixture of two products result which includes the {eta}{sup 1}(S)-DBT complex (CO){sub 5}Mo({eta}{sup 1}(S)-DBT) 4a and the unexpected {pi}-complex (CO){sub 3}Mo({eta}{sup 6}-DBT) 4b as detected by {sup 1}H NMR. The liability of the {eta}{sup 1}(S)-T{sup *} ligands is illustrated by the rapid displacement of DBT in the complex (CO){sub 5}W({eta}{sup 1}(S)-DBT) (1) by THF, and also in the complexes (CO){sub 5}Cr({eta}{sup 1}(S)-DBT) (5) and CpMn(CO){sub 2}({eta}{sup 1}(S)-DBT) (9) by CO (1 atm) at room temperature. Complexes 1-11 have been characterized spectroscopically ({sup 1}H NMR, IR) and when possible isolated as analytically pure solids (elemental analysis, EIMS). Single crystal, X-ray structural determinations are reported for (CO){sub 5}W({eta}{sup 1}(S)-DBT) and Cp(CO){sub 2}Mn({eta}{sup 1}(S)-DBT).

  6. Trace element (Al, As, B, Ba, Cr, Mo, Ni, Se, Sr, Tl, U and V) distribution and seasonality in compartments of the seagrass Cymodocea nodosa.

    PubMed

    Malea, Paraskevi; Kevrekidis, Theodoros

    2013-10-01

    Novel information on the biological fate of trace elements in seagrass ecosystems is provided. Al, As, B, Ba, Cr, Mo, Ni, Se, Sr, Tl, U and V concentrations in five compartments (blades, sheaths, vertical rhizomes, main axis plus additional branches, roots) of the seagrass Cymodocea nodosa, as well as in seawater and sediments from the Thessaloniki Gulf, Greece were determined monthly. Uni- and multivariate data analyses were applied. Leaf compartments and roots displayed higher Al, Mo, Ni and Se annual mean concentrations than rhizomes, B was highly accumulated in blades and Cr in sheaths; As, Ba, Sr and Tl contents did not significantly vary among plant compartments. A review summarizing reported element concentrations in seagrasses has revealed that C. nodosa sheaths display a high Cr accumulation capacity. Most element concentrations in blades increased in early mid-summer and early autumn with blade size and age, while those in sheaths peaked in late spring-early summer and autumn when sheath size was the lowest; elevated element concentrations in seawater in late spring and early-mid autumn, possibly as a result of elevated rainfall and associated run-off from the land, may have also contributed to the observed variability. Element concentrations in rhizomes and roots generally displayed a temporary increase in late autumn, which was concurrent with high rainfall, low wind speed associated with reduced hydrodynamism, and elevated sediment element levels. The bioaccumulation factor based on element concentrations in seagrass compartments and sediments was lower than 1 except for B, Ba, Mo, Se and Sr in all compartments, Cr in sheaths and U in roots. Blade V concentration positively correlated with sediment V concentration, suggesting that C. nodosa could be regarded as a bioindicator for V. Our findings can contribute to the design of biomonitoring programs and the development of predictive models for rational management of seagrass meadows. PMID:23838054

  7. Advanced rotor forgings for high-temperature steam turbines. Volume 2. Mechanical property evaluation. Final report. [CrMoV steels

    SciTech Connect

    Swaminathan, V.P.; Landes, J.D.

    1986-05-01

    Three advanced steel-melting processes - low-sulfur vacuum silicon deoxidation, electroslag remelting, and vacuum carbon deoxidation (VCD) - were applied to produce three CrMoV (ASTM A470, Class 8) steel forgings for steam turbine application. Ingots weighing about 100 t each were produced using these three processes, and rotors were forged with final weights of about 30 t each. Compared to the conventionally produced forgings, the advanced technology forgings show better tensile ductility and better uniformity along the radial and longitudinal directions. Charpy upper-shelf energy shows about 40% improvement, and no temper embrittlement was found using step-cooled and isothermal-aging treatments. Significant improvement in fracture toughness (K/sub IC/ and J/sub IC/) is realized for these forgings. Low-cycle fatigue life is better at high temperatures because of the absence of nonmetallic inclusions. Creep strength shows slight improvement. However, creep ductility is improved, probably because of low residual elements. The VCD forgings show excellent creep ductility, even with long lives. Both the toughness and creep properties are equal to or better than those of oil-quenched rotors produced by European practices. These improvements are attributed to cleaner steel, better control of ingot solidification, low residual elements (especially very low sulfur content), and the associated reduction of nonmetallic inclusions. These three rotors have been placed in service in three operating power plants in units rated at 520 MW each. Volume 1 of this report covers ingot and forging production, and volume 2 covers mechanical property evaluation. 40 refs., 84 figs., 15 tabs.

  8. Comparison of Ceramic, Metal and Polymer Crevice Formers on the Crevice Corrosopn Behavior of Ni-CR-Mo Alloy C22

    SciTech Connect

    X. Shan; J.H. Payer

    2006-05-08

    A necessary condition for crevice corrosion is that a crevice former create a sufficiently tight, restricted geometry on the metal surface to support the development of critical crevice chemistry. Crevice corrosion is affected by the crevice geometry (tightness) and the properties of the crevice former. The objective of this study is to determine the effect of the crevice former material on the evolution of localized corrosion-damage. A standard crevice corrosion test method is modified by (a) the use of ceramic, metal or polymer materials as the crevice former and (b) the variation of size and shape of the crevice. This study focuses on the post initiation stage of crevice corrosion and addresses factors that may limit the initiation of localized corrosion and also slow or stop the continued propagation of corrosion. Controlled crevice corrosion tests are performed under aggressive, accelerated conditions on Ni-Cr-Mo alloy C-22 and other alloys for comparison. Multiple techniques are used to examine the crevice corrosion damage evolution. Current measurements during the test provide a direct measure of the corrosion rate and indicate the initiation and any stifling or arrest. The localized corrosion is found to be stifled or arrested under several test conditions. The corrosion damage volume and profile are quantitatively measured with optical and SEM 3D reconstruction methods. Analysis by SEMIEDS, XPS and AES show that the corrosion products within the damaged crevice area are enriched in W, Mo, 0 , while being depleted in Cr, Ni, Fe. The results on C-22, SS3 16 and other alloys show that a PTFE tape covered ceramic was the most active crevice former, solid polymer crevice formers (PTFE or Kel-F) are less active, while no distinguishable crevice corrosion was observed with a ceramic material only as the crevice former in direct contact with the metal. The affects are important to the determination of the penetration rate and extent of corrosion damage by

  9. Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

    NASA Astrophysics Data System (ADS)

    T, Hino; K, Yamaguchi; Y, Yamauchi; Y, Hirohata; K, Tsuzuki; Y, Kusama

    2005-04-01

    Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study, deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus. After a ferritic steel sample was irradiated by 1.7 keV D+ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fluence dependence of retained amount of deuterium was measured by changing the ion fluence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention. Ferritic steel sample was

  10. Effects of thermal aging and neutron irradiation on the mechanical properties of stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1991-01-01

    Stainless steel weld overlay cladding was fabricated using the three-wire, series-arc method. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens. Since irradiation of the stainless steel cladding to 5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was conducted at 288{degrees}C for 1605 h, tensile, Charpy V-notch (CVN), precracked Charpy V-notch (PCVN), and compact fracture toughness specimens were thermally aged at 288{degrees}C for 1605 h. Additional specimens are being aged to 20,000 and 50,000 h. Thermal aging of three-wire, series-arc stainless steel weld overlay cladding at 288{degrees}C for 1604 h resulted in appreciable decrease (16%) in the CVN upper-shelf energy, but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect, following neutron irradiation at 288{degrees}C to a fluence of 5 {times} 10{sup 19} neutrons/cm{sup 2} (>MeV), was a 22% reduction in the CVN upper-shelf energy and a 29{degrees}C shift at the 41-J level. The effect of thermal aging on tensile properties was very small or negligible. However, the combined effect after neutron irradiation was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) and no apparent change in ultimate strength and total elongation. Also, neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging. However, irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimen become available.

  11. Effects of thermal aging and neutron irradiation on the mechanical properties of stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1991-12-31

    Stainless steel weld overlay cladding was fabricated using the three-wire, series-arc method. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens. Since irradiation of the stainless steel cladding to 5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was conducted at 288{degrees}C for 1605 h, tensile, Charpy V-notch (CVN), precracked Charpy V-notch (PCVN), and compact fracture toughness specimens were thermally aged at 288{degrees}C for 1605 h. Additional specimens are being aged to 20,000 and 50,000 h. Thermal aging of three-wire, series-arc stainless steel weld overlay cladding at 288{degrees}C for 1604 h resulted in appreciable decrease (16%) in the CVN upper-shelf energy, but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect, following neutron irradiation at 288{degrees}C to a fluence of 5 {times} 10{sup 19} neutrons/cm{sup 2} (>MeV), was a 22% reduction in the CVN upper-shelf energy and a 29{degrees}C shift at the 41-J level. The effect of thermal aging on tensile properties was very small or negligible. However, the combined effect after neutron irradiation was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) and no apparent change in ultimate strength and total elongation. Also, neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging. However, irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimen become available.

  12. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    SciTech Connect

    Ellen M. Rabenberg; Brian J. Jaques; Bulent H. Sencer; Frank A. Garner; Paula D. Freyer; Taira Okita; Darryl P. Butt

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  13. Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel

    SciTech Connect

    Onchi, T.; Hide, K.; Mayuzumi, M.; Hoshiya, T.

    2000-05-01

    Austenitic stainless steels (SS) have been used as core component materials for light water reactors. As reactors age, however, the material tends to suffer from degradation primarily resulting from irradiation-assisted stress corrosion cracking (IASCC) as well as intergranular stress corrosion cracking (IGSCC). Neutron-irradiated, thermally sensitized Type 304 (UNS S30400) stainless steels (SS) were examined by slow strain rate (SSR) stress corrosion cracking (SCC) tests in 290 C water of 0.2 ppm dissolved oxygen concentration (DO) and by SSR tensile tests in 290 C inert gas environment. Neutron fluences ranged from 4 x 10{sup 22} n/m{sup 2} to 3 x 10{sup 25} n/m{sup 2} (energy [E] > 1 MeV). percent intergranular (%IG) cracking, which has been used as an intergranular (IG) cracking susceptibility indicator in the SSR SCC tests, changes anomalously with neutron fluence in spite of the strain-to-failure rate decreasing with an increase of neutron fluence. Apparently, %IG is a misleading indicator for the irradiated, thermally sensitized Type 304 SS and for the irradiated, nonsensitized SS when IG cracking susceptibility is compared at different neutron fluences, test temperatures, DO, and strain rates. These test parameters may affect deformation and fracture behaviors of the irradiated SS during the SSR SCC tests, resulting in changing %IG, which is given by the ratio of the total IG cracking area to the entire fracture surface area. It is suggested that strain-to-IG crack initiation for the irradiated, thermally sensitized SS and for the irradiated, nonsensitized SS is the alternative indicator in the SSR SCC tests. An engineering expedient to determine the IG crack initiation strain is given by a deviating point on superposed stress-strain curves in inert gas and in oxygenated water. The strain-to-IG crack initiation becomes smaller with an increase of neutron fluence and DO. The SSR tensile tests in inert gas are needed to obtain strain-to-IG crack initiation in

  14. Advanced neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    NASA Astrophysics Data System (ADS)

    Jang, Si Young

    A heavily filtered fast neutron irradiation system (FNIS) was developed for a variety of applications, including the study of long-term health effects of fast neutrons by evaluating the biological mechanisms of damage in cultured cells and living animals such as rats or mice. This irradiation system includes an exposure cave made with a lead-bismuth alloy, a cave positioning system, a gamma and neutron monitoring system, a sample transfer system, and interchangeable filters. This system was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). By increasing the thickness of the lead-bismuth alloy, the neutron spectra were shifted into lower energies by the scattering interactions of fast neutrons with the alloy. It is possible, therefore, by changing the alloy thickness, to produce distinctly different dose weighted neutron spectra inside the exposure cave of the FNIS. The calculated neutron spectra showed close agreement with the results of activation foil measurements, unfolded by SAND-II close to the cell window. However, there was a considerable less agreement for locations far away from the cell window. Even though the magnitude of values such as neutron flux and tissue kerma rates in air differed, the weighted average neutron energies showed close agreement between the MCNP and SAND-II since the normalized neutron spectra were in a good agreement each other. A paired ion chamber system was constructed, one with a tissue equivalent plastic (A-150) and propane gas for total dose monitoring, and another with graphite and argon for photon dose monitoring. Using the pair of detectors, the neutron to gamma ratio can be inferred. With the 20 cm-thick FNIS, the absorbed dose rates of neutrons measured with the paired ion chamber method and calculated with the SAND-II results were 13.7 +/- 0.02 Gy/min and 15.5 Gy/min, respectively. The absorbed dose rate of photons and the gamma contribution to total dose were 6.7 x 10

  15. Tritium release from neutron irradiated beryllium: Kinetics, long-time annealing and effect or crack formation

    SciTech Connect

    Scaffidi-Argentina, F.; Werle, H.

    1995-09-01

    Since beryllium is considered as one of the best neutron multiplier materials in the blanket of the next generation fusion reactors, several studies have been started to evaluate its behaviour under irradiation during both operating and accidental conditions. Based on safety considerations, tritium produced in beryllium during neutron irradiation represents one important issue, therefore it is necessary to investigate tritium transport processes by using a comprehensive mathematical model and comparing its predictions with well characterized experimental tests. Because of the difficulties in extrapolating the short-time tritium release tests to a longer time scale, also long-time annealing experiments with beryllium samples from the SIBELIUS irradiation. have been carried out at the Forschungszentrum Karlsruhe. Samples were annealed up to 12 months at temperatures up to 650{degrees}C. The inventory after annealing was determined by heating the samples up to 1050{degrees}C with a He+0.1 vo1% H{sub 2} purge gas. Furthermore, in order to investigate the likely effects of cracks formation eventually causing a faster tritium release from beryllium, the behaviour of samples irradiated at low temperature (40-50{degrees}C) but up to very high fast neutron fluences (0.8-3.9{center_dot}10{sup 22} cm{sup -2}, E{sub n}{ge}1 MeV) in the BR2 reactor has been investigated. Tritium was released by heating the beryllium samples up to 1050{degrees}C and purging them with He+0.1 vo1% H{sub 2}. Tritium release from high-irradiated beryllium samples showed a much faster kinetics than from the low-irradiated ones, probably because of crack formation caused by thermal stresses in the brittle material and/or by helium bubbles migration. The obtained experimental data have been compared with predictions of the code ANFIBE with the goal to better understand the physical mechanisms governing tritium behaviour in beryllium and to assess the prediction capabilities of the code.

  16. UNDERSTANDING DAMAGE MECHANISMS IN FERRITIC/MARTENSITIC STEELS

    SciTech Connect

    Swindeman, R.W.; Maziasz, P.J.; Swindeman, M.J.

    2003-04-22

    Advanced ferritic/martensitic steels are being used extensively in fossil energy applications. New steels such as 2 1/4Cr-W-V (T23, T24), 3Cr-W-V, 9Cr-Mo-V (T91), 7Cr-W-V, 9Cr-W-V (T92 and T911), and 12Cr-W-V (T122, SAVE 12, and NF12) are examples of tubing being used in boilers and heat recovery steam generators (1). Other products for these new steels include piping, plates, and forgings. There is concern about the high-temperature performance of the advanced steels for several reasons. First, they exhibit a higher sensitivity to temperature than the 300 series stainless steels that they often replace. Second, they tend to be metallurgically unstable and undergo significant degradation at service temperatures in the creep range. Third, the experience base is limited in regard to duration. Fourth, they will be used for thick-section, high-pressure components that require high levels of integrity. To better understand the potential limitations of these steels, damage models are being developed that consider metallurgical factors as well as mechanical performance factors. Grade 91 steel was chosen as representative of these steels for evaluation of cumulative damage models since laboratory and service exposures of grade 91 exceed 100,000 hours.

  17. Laser excited novel near-infrared photoluminescence bands in fast neutron-irradiated MgO·nAl2O3

    NASA Astrophysics Data System (ADS)

    Rahman, Abu Zayed Mohammad Saliqur; Haseeb, A. S. M. A.; Xu, Qiu; Evslin, Jarah; Cinausero, Marco

    2016-08-01

    New near-infrared photoluminescence bands were observed in neutron-irradiated spinel single crystal upon excitation by a 532 nm laser. The surface morphology of the unirradiated and fast neutron-irradiated samples was investigated using atomic force microscopy and scanning probe microscopy. Fast neutron-irradiated samples show a strong emission peak at 1685 nm along with weak bands at 1065 and 2365 nm. The temperature dependence of the photoluminescence intensity was also measured. At lower temperatures, the dominant peak at 1685 nm shifts toward lower energy whereas the other peaks remain fixed. Activation energies of luminescence quenching were estimated to be 5.7 and 54.6 meV for the lower and higher temperature regions respectively.

  18. Effects of chromium and nitrogen content on the microstructures and mechanical properties of as-cast Co-Cr-Mo alloys for dental applications.

    PubMed

    Yoda, Keita; Suyalatu; Takaichi, Atsushi; Nomura, Naoyuki; Tsutsumi, Yusuke; Doi, Hisashi; Kurosu, Shingo; Chiba, Akihiko; Igarashi, Yoshimasa; Hanawa, Takao

    2012-07-01

    The microstructure and mechanical properties of as-cast Co-(20-33)Cr-5Mo-N alloys were investigated to develop ductile Co-Cr-Mo alloys without Ni addition for dental applications that satisfy the requirements of the type 5 criteria in ISO 22674. The effects of the Cr and N contents on the microstructure and mechanical properties are discussed. The microstructures were evaluated using scanning electron microscopy with energy-dispersive X-ray spectroscopy (EDS), X-ray diffractometry (XRD), and electron back-scattered diffraction pattern analysis. The mechanical properties were evaluated using tensile testing. The proof strength and elongation of N-containing 33Cr satisfied the type 5 criteria in ISO 22674. ε-phase with striations was formed in the N-free (20-29)Cr alloys, while there was slight formation of ε-phase in the N-containing (20-29)Cr alloys, which disappeared in N-containing 33Cr. The lattice parameter of the γ-phase increased with increasing Cr content (i.e. N content) in the N-containing alloys, although the lattice parameter remained almost the same in the N-free alloys because of the small atomic radius difference between Co and Cr. Compositional analyses by EDS and XRD revealed that in the N-containing alloys Cr and Mo were concentrated in the cell boundary, which became enriched in N, stabilizing the γ-phase. The mechanical properties of the N-free alloys were independent of the Cr content and showed low strength and limited elongation. Strain-induced martensite was formed in all the N-free alloys after tensile testing. On the other hand, the proof strength, ultimate tensile strength, and elongation of the N-containing alloys increased with increasing Cr content (i.e. N content). Since formation of ε-phase after tensile testing was confirmed in the N-containing alloys the deformation mechanism may change from strain-induced martensite transformation to another form, such as twinning or dislocation slip, as the N content increases. Thus the N

  19. Characterization of neutron-irradiated HT-UPS steel by high-energy X-ray diffraction microscopy

    NASA Astrophysics Data System (ADS)

    Zhang, Xuan; Park, Jun-Sang; Almer, Jonathan; Li, Meimei

    2016-04-01

    This paper presents the first measurement of neutron-irradiated microstructure using far-field high-energy X-ray diffraction microscopy (FF-HEDM) in a high-temperature ultrafine-precipitate-strengthened (HT-UPS) austenitic stainless steel. Grain center of mass, grain size distribution, crystallographic orientation (texture), diffraction spot broadening and lattice constant distributions of individual grains were obtained for samples in three different conditions: non-irradiated, neutron-irradiated (3dpa/500 °C), and irradiated + annealed (3dpa/500 °C + 600 °C/1 h). It was found that irradiation caused significant increase in grain-level diffraction spot broadening, modified the texture, reduced the grain-averaged lattice constant, but had nearly no effect on the average grain size and grain size distribution, as well as the grain size-dependent lattice constant variations. Post-irradiation annealing largely reversed the irradiation effects on texture and average lattice constant, but inadequately restored the microstrain.

  20. Neutron irradiation and frequency effects on the electrical conductivity of nanocrystalline silicon carbide (3C-SiC)

    NASA Astrophysics Data System (ADS)

    Huseynov, Elchin

    2016-09-01

    In this present work nanocrystalline silicon carbide (3C-SiC) has been irradiated with neutron flux (∼ 2 ×1013 ncm-2s-1) up to 20 hours at different periods. Electrical conductivity of nanocrystalline 3C-SiC particles (∼18 nm) is comparatively analyzed before and after neutron irradiation. The frequency dependencies of electrical conductivity of 3C-SiC nanoparticles is reviewed at 100 K-400 K temperature range before and after irradiation. The measurements were carried out at 0.1 Hz-2.5 MHz frequency ranges and at different temperatures. Radiation-induced conductivity (RIC) was observed in the nanocrystalline 3C-SiC particles after neutron irradiation and this conductivity study as a function of frequency are presented. The type of conductivity has been defined based on the interdependence between real and imaginary parts of electrical conductivity function. Based on the obtained results the mechanism behind the electrical conductivity of nanocrystalline 3C-SiC particles is explained in detail.

  1. Effects of neutron irradiation on pinning force scaling in state-of-the-art Nb3Sn wires

    NASA Astrophysics Data System (ADS)

    Baumgartner, T.; Eisterer, M.; Weber, H. W.; Flükiger, R.; Scheuerlein, C.; Bottura, L.

    2014-01-01

    We present an extensive irradiation study involving five state-of-the-art Nb3Sn wires which were subjected to sequential neutron irradiation up to a fast neutron fluence of 1.6 × 1022 m-2 (E > 0.1 MeV). The volume pinning force of short wire samples was assessed in the temperature range from 4.2 to 15 K in applied fields of up to 7 T by means of SQUID magnetometry in the unirradiated state and after each irradiation step. Pinning force scaling computations revealed that the exponents in the pinning force function differ significantly from those expected for pure grain boundary pinning, and that fast neutron irradiation causes a substantial change in the functional dependence of the volume pinning force. A model is presented, which describes the pinning force function of irradiated wires using a two-component ansatz involving a point-pinning contribution stemming from radiation induced pinning centers. The dependence of this point-pinning contribution on fast neutron fluence appears to be a universal function for all examined wire types.

  2. 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.

  3. RhG-CSF improves radiation-induced myelosuppression and survival in the canine exposed to fission neutron irradiation.

    PubMed

    Yu, Zu-Yin; Li, Ming; Han, A-Ru-Na; Xing, Shuang; Ou, Hong-Ling; Xiong, Guo-Lin; Xie, Ling; Zhao, Yan-Fang; Xiao, He; Shan, Ya-Jun; Zhao, Zhen-Hu; Liu, Xiao-Lan; Cong, Yu-Wen; Luo, Qing-Liang

    2011-01-01

    Fission-neutron radiation damage is hard to treat due to its critical injuries to hematopoietic and gastrointestinal systems, and so far few data are available on the therapeutic measures for neutron-radiation syndrome. This study was designed to test the effects of recombinant human granulocyte colony-stimulating factor (rhG-CSF) in dogs which had received 2.3 Gy mixed fission-neutron-γ irradiation with a high ratio of neutrons (~90%). Following irradiation, rhG-CSF treatment induced 100% survival versus 60% in controls. Only two of five rhG-CSF-treated dogs experienced leukopenia (white blood cells [WBC] count < 1.0 × 10(9)/L) and neutropenia (neutrophil [ANC] count < 0.5 × 10(9)/L), whereas all irradiated controls displayed a profound period of leukopenia and neutropenia. Furthermore, administration of rhG-CSF significantly delayed the onset of leukopenia and reduced the duration of leucopenia as compared with controls. In addition, individual dogs in the rhG-CSF-treated group exhibited evident differences in rhG-CSF responsiveness after neutron-irradiation. Finally, histopathological evaluation of the surviving dogs revealed that the incidence and severity of bone marrow, thymus and spleen damage decreased in rhG-CSF-treated dogs as compared with surviving controls. Thus, these results demonstrated that rhG-CSF administration enhanced recovery of myelopoiesis and survival after neutron-irradiation. PMID:21785235

  4. 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.

  5. 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.

  6. Comminuting irradiated ferritic steel

    DOEpatents

    Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

    1985-01-01

    Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

  7. Contribution To Degradation Study, Behavior Of Unsaturated Polyester Resin Under Neutron Irradiation

    NASA Astrophysics Data System (ADS)

    Abellache, D.; Lounis, A.; Taïbi, K.

    2010-01-01

    Applications of unsaturated polyester thermosetting resins are numerous in construction sector, in transport, electric spare parts manufactures, consumer goods, and anticorrosive materials. This survey reports the effect of thermosetting polymer degradation (unsaturated polyester): degradation by neutrons irradiation. In order to evaluate the deterioration of our material, some comparative characterizations have been done between standard samples and damaged ones. Scanning electron microscopy (SEM), ultrasonic scanning, hardness test (Shore D) are the techniques which have been used. The exposure to a neutrons flux is carried out in the column of the nuclear research reactor of Draria (Algiers-Algeria). The energetic profile of the incidental fluxes is constituted of fast neutrons (ΦR = 3.1012n.cm-2.s-1, E = 2 Mev) of thermal neutrons (ΦTH = 1013n.cm-2.s-1; E = 0.025 ev) and epithermal neutrons (Φepi = 7.1011 n.cm-2.s-1; E>4,9 ev). The received dose flow is 0,4 Kgy. We notice only a few scientific investigations can be found in this field. In comparison with the standard sample (no exposed) it is shown that the damage degree is an increasing process with the exposure. Concerning the description of irradiation effects on polymers, we can advance that several reactions are in competition : reticulation, chain break, and oxidation by radical mechanism. In our case the incidental particle of high energy fast neutrons whose energy is greater or equal to 2 Mev, is braked by the target with a nuclear shock during which the incidental particle transmits a part of its energy to an atom. If the energy transfer is sufficient, the nuclear shock permits to drive out an atom of its site the latter will return positioning interstitially, the energy that we used oversteps probably the energy threshold (displacement energy). This fast neutrons collision with target cores proceeds to an indirect ionization by the preliminary creation of excited secondary species that will

  8. Contribution To Degradation Study, Behavior Of Unsaturated Polyester Resin Under Neutron Irradiation

    SciTech Connect

    Abellache, D.; Lounis, A.; Taiebi, K.

    2010-01-05

    Applications of unsaturated polyester thermosetting resins are numerous in construction sector, in transport, electric spare parts manufactures, consumer goods, and anticorrosive materials. This survey reports the effect of thermosetting polymer degradation (unsaturated polyester): degradation by neutrons irradiation. In order to evaluate the deterioration of our material, some comparative characterizations have been done between standard samples and damaged ones. Scanning electron microscopy (SEM), ultrasonic scanning, hardness test (Shore D) are the techniques which have been used. The exposure to a neutrons flux is carried out in the column of the nuclear research reactor of Draria (Algiers-Algeria). The energetic profile of the incidental fluxes is constituted of fast neutrons (PHI{sub R} = 3.10{sup 12} n.cm{sup -2}.s{sup -1}, E = 2 Mev) of thermal neutrons (PHI{sub TH} = 10{sup 13} n.cm{sup -2}.s{sup -1}; E = 0.025 ev) and epithermal neutrons (PHI{sub epi} = 7.10{sup 11} n.cm{sup -2}.s{sup -1}; E>4,9 ev). The received dose flow is 0,4 Kgy. We notice only a few scientific investigations can be found in this field. In comparison with the standard sample (no exposed) it is shown that the damage degree is an increasing process with the exposure. Concerning the description of irradiation effects on polymers, we can advance that several reactions are in competition: reticulation, chain break, and oxidation by radical mechanism. In our case the incidental particle of high energy fast neutrons whose energy is greater or equal to 2 Mev, is braked by the target with a nuclear shock during which the incidental particle transmits a part of its energy to an atom. If the energy transfer is sufficient, the nuclear shock permits to drive out an atom of its site the latter will return positioning interstitially, the energy that we used oversteps probably the energy threshold (displacement energy). This fast neutrons collision with target cores proceeds to an indirect

  9. Atomistic study on mixed-mode fracture mechanisms of ferrite iron interacting with coherent copper and nickel nanoclusters

    NASA Astrophysics Data System (ADS)

    Al-Motasem, Ahmed Tamer; Mai, Nghia Trong; Choi, Seung Tae; Posselt, Matthias

    2016-04-01

    The effect of copper and/or nickel nanoclusters, generally formed by neutron irradiation, on fracture mechanisms of ferrite iron was investigated by using molecular statics simulation. The equilibrium configuration of nanoclusters was obtained by using a combination of an on-lattice annealing based on Metropolis Monte Carlo method and an off-lattice relaxation by molecular dynamics simulation. Residual stress distributions near the nanoclusters were also calculated, since compressive or tensile residual stresses may retard or accelerate, respectively, the propagation of a crack running into a nanocluster. One of the nanoclusters was located in front of a straight crack in ferrite iron with a body-centered cubic crystal structure. Two crystallographic directions, of which the crack plane and crack front direction are (010)[001] and (111) [ 1 bar 10 ] , were considered, representing cleavage and non-cleavage orientations in ferrite iron, respectively. Displacements corresponding to pure opening-mode and mixed-mode loadings were imposed on the boundary region and the energy minimization was performed. It was observed that the fracture mechanisms of ferrite iron under the pure opening-mode loading are strongly influenced by the presence of nanoclusters, while under the mixed-mode loading the nanoclusters have no significant effect on the crack propagation behavior of ferrite iron.

  10. XXIst Century Ferrites

    NASA Astrophysics Data System (ADS)

    Mazaleyrat, F.; Zehani, K.; Pasko, A.; Loyau, V.; LoBue, M.

    2012-05-01

    Ferrites have always been a subject of great interest from point of view of magnetic application, since the fist compass to present date. In contrast, the scientific interest for iron based magnetic oxides decreased after Ørsted discovery as they where replaced by coil as magnetizing sources. Neel discovery of ferrimagnetism boosted again interest and leads to strong developments during two decades before being of less interest. Recently, the evolution of power electronics toward higher frequency, the downsizing of ceramics microstucture to nanometer scale, the increasing price of rare-earth elements and the development of magnetocaloric materials put light again on ferrites. A review on three ferrite families is given herein: harder nanostructured Ba2+Fe12O19 magnet processed by spark plasma sintering, magnetocaloric effect associated to the spin transition reorientation of W-ferrite and low temperature spark plasma sintered Ni-Zn-Cu ferrites for high frequency power applications.

  11. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    SciTech Connect

    Franco, Manuel,

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

  12. X-ray diffraction study of BaTiO{sub 3} single crystals before and after fast-neutron irradiation

    SciTech Connect

    Stash, A. I. Ivanov, S. A.; Stefanovich, S. Yu.; Mosunov, A. V.; Boyko, V. M.; Ermakov, V. S.; Korulin, A. V.; Kalyukanov, A. I.; Isakova, N. N.

    2015-09-15

    The neutron irradiation of ferroelectrics is efficiently used to form structural states that cannot be obtained by conventional technologies. To date, the effect of neutron irradiation on the structure and properties of BaTiO{sub 3} has been studied for only ceramic materials. We have considered the influence of fast-neutron irradiation (F = 1 × 10{sup 17} cm{sup −2}) on the structure and properties of BaTiO{sub 3} single crystals for the first time. The structural changes occurring in irradiated BaTiO{sub 3} and their correlation with the behavior of dielectric and nonlinear optical characteristics are analyzed with the aid of a specially developed method for taking into account the experimental correction to diffuse scattering. Neutron irradiation to the aforementioned dose retains the polar structure of the material and only slightly changes atomic displacements. The radiationinduced structural changes occur according to the high-temperature type to form a structure similar to the cubic modification of unirradiated BaTiO{sub 3} crystal.

  13. Inverse magnetocaloric effect in Ce(Fe{sub 0.96}Ru{sub 0.04}){sub 2}: Effect of fast neutron irradiation

    SciTech Connect

    Dube, V.; Mishra, P. K.; Prajapat, C. L.; Singh, M. R.; Ravikumar, G.; Rajarajan, A. K.; Sastry, P. U.; Thakare, S. V.

    2013-02-05

    We have shown the effect of fast neutron irradiation on the magnetic phase transition and magnetocaloric effect (MCE) in a doped Ce(Fe{sub 0.96}Ru{sub 0.04}){sub 2}, intermettalic. We show that this leads to suppression of MCE and a to a disordered ferromagnetic phase.

  14. Temperature effect on characteristics of void population formed in the austenitic steel under neutron irradiation up to high damage dose

    NASA Astrophysics Data System (ADS)

    Kozlov, A. V.; Portnykh, I. A.; Skryabin, L. A.; Kinev, E. A.

    2002-12-01

    Radiation-induced porosity in fuel pin cladding of the BN-600 reactor fabricated of cold-worked austenitic steel 16Cr-15Ni-2Mo-2Mn irradiated to different damage dose 20-90 dpa at 410-600 °C has been examined by transmission electron microscopy. Formation and growth of various types of voids were shown to occur according to their both duration and mechanism of nucleation. Dependencies of average diameters and concentration of all void types on neutron irradiation damage dose were plotted for various temperature ranges. The change of void population with increasing dose at various temperature ranges was analyzed based on point defect kinetic. The contribution of different types of voids to swelling was examined.

  15. Microstructural evolution of NF709 (20Cr–25Ni–1.5MoNbTiN) under neutron irradiation

    DOE PAGESBeta

    Kim, Byoungkoo; Tan, Lizhen; Xu, C.; Yang, Yong; Zhang, Xuan; Li, Meimei

    2015-12-30

    In this study, because of its superior creep and corrosion resistance as compared with general austenitic stainless steels, NF709 has emerged as a candidate structural material for advanced nuclear reactors. To obtain fundamental information about the radiation resistance of this material, this study examined the microstructural evolution of NF709 subjected to neutron irradiation to 3 displacements per atom at 500 °C. Transmission electron microscopy, scanning electron microscopy, and high-energy x-ray diffraction were employed to characterize radiation-induced segregation, Frank loops, voids, as well as the formation and reduction of precipitates. Radiation hardening of ~76% was estimated by nanoindentation, approximately consistent withmore » the calculation according to the dispersed barrier-hardening model, suggesting Frank loops as the primary hardening source.« less

  16. Degradation of Nylon 6,6 Fire-Suppression Casing from Plutonium Glove Boxes Under Alpha and Neutron Irradiation

    DOE PAGESBeta

    Millsap, Donald W.; Cournoyer, Michael E.; Landsberger, Sheldon; Tesmer, Joseph R.; Wang, Matthew Y.

    2015-04-23

    Nylon 6,6 tensile specimens, conforming to the casing for self-contained fire extinguisher systems, have been irradiated using both an accelerator He++ ion beam and a 5-Ci PuBe neutron source to model the radiation damage these systems would likely incur over a lifetime of operation within glove boxes. Following irradiation, these samples were mechanically tested using standard practices as described in ASTM D638. The results of the He++ study indicate that the tensile strength of the nylon specimens undergoes some slight (<10%) degradation while other properties of the samples, such as elongation and tangent modulus, appear to fluctuate with increasing dosemore » levels. The He++-irradiated specimens also have a noticeable level of discoloration corresponding to increasing levels of dose. The neutron-irradiated samples show a higher degree of mechanical degradation than the He++-irradiated samples.« less

  17. Effects of neutron irradiation on dimensional stability and on mechanical properties of SiC/SiC composites

    SciTech Connect

    Youngblood, G.E.; Henager, C.H. Jr.; Senor, J.

    1995-04-01

    The objective of this work is to assess the development and the performance of continuous fiber SiC{sub f}/SiC composites as a structural material for advanced fusion reactor application. The dimensional stability and some mechanical properties of two similar 2D 0-90{degree} weave SiC{sub f}/SiC composites made with Nacalon{trademark} ceramic-grade fiber were characterized and compared after neutron irradiation to those properties for {beta}-SiC. The major difference between these two composites was that one had a thin (150 nm) and the other a thick (1000 nm) graphite interface layer. The irradiation conditions consisted of relatively high doses (4.3 to 26 dpa-SiC) at high temperature (430-1200{degree}C).

  18. Microstructural evolution of NF709 (20Cr-25Ni-1.5MoNbTiN) under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kim, B. K.; Tan, L.; Xu, C.; Yang, Y.; Zhang, X.; Li, M.

    2016-03-01

    Because of its superior creep and corrosion resistance as compared with general austenitic stainless steels, NF709 has emerged as a candidate structural material for advanced nuclear reactors. To obtain fundamental information about the radiation resistance of this material, this study examined the microstructural evolution of NF709 subjected to neutron irradiation to 3 displacements per atom at 500 °C. Transmission electron microscopy, scanning electron microscopy, and high-energy x-ray diffraction were employed to characterize radiation-induced segregation, Frank loops, voids, as well as the formation and reduction of precipitates. Radiation hardening of ∼76% was estimated by nanoindentation, approximately consistent with the calculation according to the dispersed barrier-hardening model, suggesting Frank loops as the primary hardening source.

  19. Reconstitution and Upgrade of the Thermal Neutron Irradiation Facility in the Basement Medical Room of the MIT Research Reactor

    SciTech Connect

    Harling, Otto, K.; Riley, Kent, J.; Binns, Peter J.

    2004-12-31

    The M-011 thermal neutron beam has been reconstituted and upgraded to provide a high intensity and high quality facility for preclinical and certain clinical studies. Intensities of thermal neutrons in the beam range from 5.0-8.5 x 109 n cm-2 s-1. Beam contamination is at a low level where it has no practical influence on beam performance. New computer controlled dose and beam monitoring systems have been implemented which assure precise dose delivery and redundant safety interlocks. An additional beam shutter and massive shielding in the back of the medical room have been added which significantly reduce room background and now permit staff entry without the necessity for lowering the reactor power. This system is needed for BNCT research by the MIT group as well as other US groups. This need became acute with the closure of the BMRR which previously had the only high quality thermal neutron irradiation facility for BNCT in the USA.

  20. Degradation of Nylon 6,6 Fire-Suppression Casing from Plutonium Glove Boxes Under Alpha and Neutron Irradiation

    SciTech Connect

    Millsap, Donald W.; Cournoyer, Michael E.; Landsberger, Sheldon; Tesmer, Joseph R.; Wang, Matthew Y.

    2015-04-23

    Nylon 6,6 tensile specimens, conforming to the casing for self-contained fire extinguisher systems, have been irradiated using both an accelerator He++ ion beam and a 5-Ci PuBe neutron source to model the radiation damage these systems would likely incur over a lifetime of operation within glove boxes. Following irradiation, these samples were mechanically tested using standard practices as described in ASTM D638. The results of the He++ study indicate that the tensile strength of the nylon specimens undergoes some slight (<10%) degradation while other properties of the samples, such as elongation and tangent modulus, appear to fluctuate with increasing dose levels. The He++-irradiated specimens also have a noticeable level of discoloration corresponding to increasing levels of dose. The neutron-irradiated samples show a higher degree of mechanical degradation than the He++-irradiated samples.

  1. Characterization of phosphorus segregation in neutron-irradiated pressure vessel steels by atom probe field ion microscopy

    SciTech Connect

    Miller, M.K.; Jayaram, R.; Russell, K.F.

    1995-04-01

    An atom probe field ion microscopy characterization of A533B and Russian VVER 440 and 1000 pressure vessel steels has been performed to determine the phosphorus coverage of grain and lath boundaries. Field ion micrographs of grain and lath boundaries have revealed that they are decorated with a semi-continuous film of discrete brightly-imaging precipitates that were identified as molybdenum carbonitride precipitates. In addition, extremely high phosphorus levels were measured at the boundaries. The phosphorus segregation was found to be confined to an extremely narrow region indicative of monolayer-type segregation. The phosphorus coverages determined from the atom probe results of the unirradiated materials were in excellent agreement with predictions based on McLean`s equilibrium model of grain boundary segregation. The boundary phosphorus coverage of a neutron-irradiated weld material was significantly higher than that observed in the unirradiated material.

  2. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    NASA Astrophysics Data System (ADS)

    Rabenberg, Ellen M.; Jaques, Brian J.; Sencer, Bulent H.; Garner, Frank A.; Freyer, Paula D.; Okita, Taira; Butt, Darryl P.

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. An effective tensile strain hardening exponent was also obtained from the data which shows a relative decrease in ductility of steel with increased irradiation damage. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  3. Microstructural evolution of NF709 (20Cr–25Ni–1.5MoNbTiN) under neutron irradiation

    SciTech Connect

    Kim, Byoungkoo; Tan, Lizhen; Xu, C.; Yang, Yong; Zhang, Xuan; Li, Meimei

    2015-12-30

    In this study, because of its superior creep and corrosion resistance as compared with general austenitic stainless steels, NF709 has emerged as a candidate structural material for advanced nuclear reactors. To obtain fundamental information about the radiation resistance of this material, this study examined the microstructural evolution of NF709 subjected to neutron irradiation to 3 displacements per atom at 500 °C. Transmission electron microscopy, scanning electron microscopy, and high-energy x-ray diffraction were employed to characterize radiation-induced segregation, Frank loops, voids, as well as the formation and reduction of precipitates. Radiation hardening of ~76% was estimated by nanoindentation, approximately consistent with the calculation according to the dispersed barrier-hardening model, suggesting Frank loops as the primary hardening source.

  4. Critical current densities in neutron irradiated Tl 2Ca 2Ba 2Cu 3O 10 single crystals

    NASA Astrophysics Data System (ADS)

    Brandstätter, G.; Sauerzopf, F. M.; Weber, H. W.; Aghaei, A.; Schwarzmann, F.

    1994-12-01

    A Tl 2Ca 2Ba 2Cu 3O 10 single crystal with a transition temperature of 117.5 K was subjected to fast neutron irradiation to fluences of 2·10 21, 4·10 21, 8·10 21, and 1.6·10 22 m 2 (E>0.1 MeV). The superconducting transition temperatures T c, the hysteresis loops and the irreversibility lines were measured before and after each irradiation step. The critical current densities J c were calculated from the magnetization loops using an anisotropic Bean model. With increasing fluence we find a decrease of T c, as observed in YBCO-123 and other high temperature superconductors, and an increase of J c. The irreversibility line is shifted to higher fields and temperatures.

  5. High-dose neutron irradiation of Hi-Nicalon Type S silicon carbide composites. Part 2: Mechanical and physical properties

    NASA Astrophysics Data System (ADS)

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wally; Snead, Lance L.

    2015-07-01

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573-1073 K. The material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. The observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  6. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

    SciTech Connect

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  7. Development and characteristics of the HANARO neutron irradiation facility for applications in the boron neutron capture therapy field

    NASA Astrophysics Data System (ADS)

    Kim, Myong-Seop; Lee, Byung-Chul; Hwang, Sung-Yul; Kim, Heonil; Jun, Byung-Jin

    2007-05-01

    The HANARO neutron irradiation facility for various applications in the boron neutron capture therapy (BNCT) field was developed, and its characteristics were investigated. In order to obtain the sufficient thermal neutron flux with a low level of contamination by fast neutrons and gamma rays, a radiation filtering method was adopted. The radiation filter was designed by using a silicon single crystal, cooled by liquid nitrogen, and a bismuth crystal. The installation of the main components of the irradiation facility and the irradiation room was finished. Neutron beam characteristics were measured by using bare and cadmium-covered gold foils and wires. The in-phantom neutron flux distribution was measured for flux mapping inside the phantom. The gamma-ray dose was determined by using TLD-700 thermoluminescence dosimeters. The thermal and fast neutron fluxes and the gamma-ray dose were calculated by using the MCNP code, and they were compared with experimental data. The thermal neutron flux and Cd ratio available at this facility were confirmed to be 1.49 × 109 n cm-2 s-1 and 152, respectively. The maximum neutron flux inside the phantom was measured to be 2.79 × 109 n cm-2 s-1 at a depth of 3 mm in the phantom. The two-dimensional in-phantom neutron flux distribution was determined, and significant neutron irradiation was observed within 20 mm from the phantom surface. The gamma-ray dose rate for the free beam condition was expected to be about 80 cGy h-1. These experimental results were reasonably well supported by calculation using the facility design code. This HANARO thermal neutron facility can be used not only for clinical trials, but also for various pre-clinical studies in the BNCT field.

  8. Development and characteristics of the HANARO neutron irradiation facility for applications in the boron neutron capture therapy field.

    PubMed

    Kim, Myong-Seop; Lee, Byung-Chul; Hwang, Sung-Yul; Kim, Heonil; Jun, Byung-Jin

    2007-05-01

    The HANARO neutron irradiation facility for various applications in the boron neutron capture therapy (BNCT) field was developed, and its characteristics were investigated. In order to obtain the sufficient thermal neutron flux with a low level of contamination by fast neutrons and gamma rays, a radiation filtering method was adopted. The radiation filter was designed by using a silicon single crystal, cooled by liquid nitrogen, and a bismuth crystal. The installation of the main components of the irradiation facility and the irradiation room was finished. Neutron beam characteristics were measured by using bare and cadmium-covered gold foils and wires. The in-phantom neutron flux distribution was measured for flux mapping inside the phantom. The gamma-ray dose was determined by using TLD-700 thermoluminescence dosimeters. The thermal and fast neutron fluxes and the gamma-ray dose were calculated by using the MCNP code, and they were compared with experimental data. The thermal neutron flux and Cd ratio available at this facility were confirmed to be 1.49 x 10(9) n cm(-2) s(-1) and 152, respectively. The maximum neutron flux inside the phantom was measured to be 2.79 x 10(9) n cm(-2) s(-1) at a depth of 3 mm in the phantom. The two-dimensional in-phantom neutron flux distribution was determined, and significant neutron irradiation was observed within 20 mm from the phantom surface. The gamma-ray dose rate for the free beam condition was expected to be about 80 cGy h(-1). These experimental results were reasonably well supported by calculation using the facility design code. This HANARO thermal neutron facility can be used not only for clinical trials, but also for various pre-clinical studies in the BNCT field. PMID:17440252

  9. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  10. Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1997-05-01

    Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect of aging and neutron irradiation at 288{degrees}C to a fluence of 5 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343{degrees}C for 20,000 h each were very small and similar to those at 288{degrees}C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288{degrees}C will be investigated as the specimens become available in 1996 and beyond.

  11. 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.

  12. Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

    SciTech Connect

    Masashi Shimada; Y. Hatano; Y. Oya; T. Oda; M. Hara; G. Cao; M. Kobayashi; M. Sokolov; H. Watanabe; B. Tyburska; Y. Ueda; P. Calderoni

    2011-09-01

    Plasma-facing components (PFCs) will be exposed to 14 MeV neutrons from deuterium-tritium (D-T) fusion reactions, and tungsten, a candidate PFC for the divertor in ITER, is expected to receive a neutron dose of 0.7 displacement per atom (dpa) by the end of operation in ITER. The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in PFCs, it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high PKA energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influence the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron irradiation. Therefore, the effort to correlate among high-energy ions, fission neutrons, and fusion neutrons is crucial for accurately estimating tritium retention under a neutron-irradiation environment. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at. % purity from A.L.M.T. Co.) were irradiated by neutron in the High Flux Isotope Reactor (HFIR), ORNL, at 50 and 300C to 0.025, 0.3, and 1.2 dpa, and the investigation of deuterium retention in neutron-irradiation was performed in the INL Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe2+, 20 MeV W2+, and 700 keV H-) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

  13. Impedance calculation for ferrite inserts

    SciTech Connect

    Breitzmann, S.C.; Lee, S.Y.; Ng, K.Y.; /Fermilab

    2005-01-01

    Passive ferrite inserts were used to compensate the space charge impedance in high intensity space charge dominated accelerators. They study the narrowband longitudinal impedance of these ferrite inserts. they find that the shunt impedance and the quality factor for ferrite inserts are inversely proportional to the imaginary part of the permeability of ferrite materials. They also provide a recipe for attaining a truly passive space charge impedance compensation and avoiding narrowband microwave instabilities.

  14. The role of nickel in radiation damage of ferritic alloys

    DOE PAGESBeta

    Osetskiy, Yury N.; Anento, Napoleon; Serra, Anna; Terentyev, Dmitry

    2014-11-26

    According to the modern theory damage evolution under neutron irradiation depends on the fraction of self interstitial atoms (SIAs) produced in the form of one-dimensionally (1-D) glissile clusters. These clusters, having a low interaction cross-section with other defects, sink mainly on grain boundaries and dislocations creating the so-called production bias. It is known empirically that addition of certain alloying elements affect many radiation effects, including swelling, however the mechanisms are unknown in many cases. In this paper we report the results of an extensive multi-technique atomistic level modeling of SIA clusters mobility in bcc Fe-Ni alloys with Ni content frommore » 0.8 to 10 at.%. We have found that Ni interacts strongly with periphery of clusters affecting their mobility. The total effect is defined by all Ni atoms interacting with the cluster at the same time and can be significant even in low-Ni alloys. Thus 1nm (37SIAs) cluster is practically immobile at T < 500K in the Fe-0.8at.% Ni alloy. Increasing cluster size and Ni content enhance cluster immobilization. Furthermore, this effect should have quite broad consequences in swelling rate, matrix damage accumulation, radiation induced hardening, etc. and the results obtained help in better understanding and prediction of radiation effects in Fe-Ni ferritic alloys.« less

  15. The role of nickel in radiation damage of ferritic alloys

    SciTech Connect

    Osetskiy, Yury N.; Anento, Napoleon; Serra, Anna; Terentyev, Dmitry

    2014-11-26

    According to the modern theory damage evolution under neutron irradiation depends on the fraction of self interstitial atoms (SIAs) produced in the form of one-dimensionally (1-D) glissile clusters. These clusters, having a low interaction cross-section with other defects, sink mainly on grain boundaries and dislocations creating the so-called production bias. It is known empirically that addition of certain alloying elements affect many radiation effects, including swelling, however the mechanisms are unknown in many cases. In this paper we report the results of an extensive multi-technique atomistic level modeling of SIA clusters mobility in bcc Fe-Ni alloys with Ni content from 0.8 to 10 at.%. We have found that Ni interacts strongly with periphery of clusters affecting their mobility. The total effect is defined by all Ni atoms interacting with the cluster at the same time and can be significant even in low-Ni alloys. Thus 1nm (37SIAs) cluster is practically immobile at T < 500K in the Fe-0.8at.% Ni alloy. Increasing cluster size and Ni content enhance cluster immobilization. Furthermore, this effect should have quite broad consequences in swelling rate, matrix damage accumulation, radiation induced hardening, etc. and the results obtained help in better understanding and prediction of radiation effects in Fe-Ni ferritic alloys.

  16. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed. PMID:12408308

  17. Ferrite logic reliability study

    NASA Technical Reports Server (NTRS)

    Baer, J. A.; Clark, C. B.

    1973-01-01

    Development and use of digital circuits called all-magnetic logic are reported. In these circuits the magnetic elements and their windings comprise the active circuit devices in the logic portion of a system. The ferrite logic device belongs to the all-magnetic class of logic circuits. The FLO device is novel in that it makes use of a dual or bimaterial ferrite composition in one physical ceramic body. This bimaterial feature, coupled with its potential for relatively high speed operation, makes it attractive for high reliability applications. (Maximum speed of operation approximately 50 kHz.)

  18. Investigation of X-ray spectral response of D-T fusion produced neutron irradiated PIPS detectors for plasma X-ray diagnostics

    NASA Astrophysics Data System (ADS)

    Vigneshwara Raja, P.; Narasimha Murty, N. V. L.; Rao, C. V. S.; Abhangi, Mitul

    2015-10-01

    This paper describes the fusion-produced neutron irradiation induced changes in the X-ray spectral response of commercially available Passivated Implanted Planar Silicon (PIPS) detectors using the accelerator based D-T generator. After 14.1 MeV neutron irradiation up to a fluence of 3.6× 1010 n/cm2, the energy resolution (i.e. FWHM) of the detectors at room temperature is found to degrade by about 3.8 times that of the pre-irradiated value. From the X-ray spectral characteristics, it has been observed that the room temperature spectral response of PIPS detectors is too poor even at low neutron fluences. Irradiation is also carried out with Am-Be neutron source for studying the effect of scattered neutrons from the reactor walls on the detector performance. Comparative studies of the damage caused by 14.1 MeV neutrons and Am-Be source produced neutrons at the same neutron fluence are carried out by analyzing the irradiated detector characteristics. The degradation in the energy resolution of the detectors is attributed to the radiation induced changes in the detector leakage current. No considerable changes in the full depletion voltage and the effective doping concentration up to the neutron fluence of 3.6× 1010 n/cm2, are observed from the measured C-V characteristics. Partial recovery of the neutron irradiated detector characteristics is discussed.

  19. Neutron-transmuted carbon-14 in neutron-irradiated GaN: Compensation of DX-like center

    SciTech Connect

    Ida, T.; Oga, T.; Kuriyama, K.; Kushida, K.; Xu, Q.; Fukutani, S.

    2013-12-04

    The transmuted-C related luminescence and net carrier concentration are studied by combining photoluminescence, liquid scintillation, and Raman scattering. GaN single crystal films grown by metalorganic-vapor-phase epitaxy are irradiated with fast and thermal neutrons at fluxes of 3.9 × 10{sup 13} cm{sup −2}s{sup −1} and 8.15 × 10{sup 13} cm{sup −2}s{sup −1}, respectively. Irradiation time is 48 hours. The calculated {sup 72}Ge and {sup 14}C concentrations are 1.24 × 10{sup 18} cm{sup −3} and 1.13 × 10{sup 18} cm{sup −3}, respectively. The transmuted {sup 14}C is detected by the liquid scintillation method to survey β-rays emitted in the process of {sup 14}C decays from {sup 14}N. Tritium ({sup 3}H) is also emitted by a (n,t) reaction of {sup 14}N due to the neutron irradiation above 4.5 MeV. Photoluminescence relating to C, DX-like center of Ge and yellow luminescence band are observed in 1000 °C annealed NTD-GaN. The free electron concentration estimated from Raman scattering is 4.97 × 10{sup 17} cm{sup −3}. This value is lower than that from the transmuted Ge concentration, suggesting the compensation due to the transmuted {sup 14}C acceptors.

  20. Changes in magnetic parameters of neutron irradiated SA 508 Cl. 3 reactor pressure vessel forging and weld surveillance specimens

    NASA Astrophysics Data System (ADS)

    Chi, Se-Hwan; Chang, Kee-Ok; Hong, Jun-Hwa; Kuk, Il-Hiun; Kim, Chong-Oh

    1999-04-01

    Irradiation-induced changes in the magnetic parameters and mechanical properties were measured and compared to explore possible correlations for reactor pressure vessel (RPV) forging and weld surveillance Charpy specimens which were irradiated to the neutron fluence of 2.3×1019n/cm2 (E>1.0 MeV) in a typical pressurized water reactor environment at 290 °C. For mechanical property parameters, Vickers microhardness, tensile and Charpy impact tests were performed and saturation magnetization (Ms), remanence (Mr), coercivity (Hc), and Barkhausen noise amplitude (BNA) were measured for magnetic parameters for both unirradiated and irradiated specimens, respectively. Results of mechanical property measurements showed an increase in yield and tensile strength, Vickers microhardness, 30 ft. lb indexed RTNDT and a decrease in Charpy upper-shelf energy irrespective of forging and weld metals. Hysteresis loops appeared to turn clockwise, resulting in an increase in Hc, and BNA appeared to decrease after irradiation. Both magnetic parameters showed viable correlations to the changes in mechanical parameters (Vickers microhardness, Charpy upper shelf energy) due to irradiation. Even limited, the present study seems to show additional possibilities for the application of this magnetic method in monitoring the mechanical parameter changes due to neutron irradiation.

  1. FISSION NEUTRON IRRADIATION EFFECT ON INTERLAMINAR SHEAR STRENGTH OF CYANATE ESTER RESIN GFRP AT RT AND 77 K

    SciTech Connect

    Nishimura, A.; Izumi, Y.; Nishijima, S.; Hemmi, T.; Koizumi, K.; Takeuchi, T.; Shikama, T.

    2010-04-08

    A glass fiber reinforced plastic (GFRP) with cyanate ester resin was fabricated and neutron irradiation tests up to 1x10{sup 22} n/m{sup 2} of fast neutron with over 0.1 MeV energy were carried out in fission reactor. The fabrication process of cyanate ester GFRP was established and a collaboration network to perform investigations on irradiation effect of superconducting magnet materials was constructed. Three kinds of samples were fabricated. The first was CTD403 GFRP made by NIFS, the second was (cyanate ester+epoxy) GFRP provided by Toshiba, and the last was CTD403 GFRP made by Toshiba. The irradiation was carried out at JRR-3 in Japan Atomic Energy Agency using Rabbit capsules.After the irradiation, short beam tests were conducted at room temperature and 77 K and interlaminar shear strength (ILSS) was evaluated. The irradiation of 1x10{sup 21} n/m{sup 2} increased ILSS a little but 1x10{sup 22} n/m{sup 2} irradiation decreased ILSS to around 50 MPa. These tendencies were observed in all three kinds of GFRPs.

  2. Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358 C

    SciTech Connect

    Cockeram, Brian V; Smith, Richard W; Leonard, Keith J; Byun, Thak Sang; Snead, Lance Lewis

    2011-01-01

    Wrought Zircaloy-2 and Zircaloy-4 were neutron irradiated at nominally 358 C in the high flux isotope reactor (HFIR) at relatively low neutron fluences between 5.8 1022 and 2.9 1025 n/m2 (E > 1 MeV). The irradiation hardening and change in microstructure were characterized following irradiation using tensile testing and examinations of microstructure using Analytical Electron Microscopy (AEM). Small increments of dose (0.0058, 0.11, 0.55, 1.08, and 2.93 1025 n/m2) were used in the range where the saturation of irradiation hardening is typically observed so that the role of microstructure evolution and hai loop formation on irradiation hardening could be correlated. An incubation dose between 5.8 1023 and 1.1 1024 n/m2 was needed for loop nucleation to occur that resulted in irradiation hardening. Increases in yield strength were consistent with previous results in this temperature regime, and as expected less irradiation hardening and lower hai loop number density values than those generally reported in literature for irradiations at 260 326 C were observed. Unlike previous lower temperature data, there is evidence in this study that the irradiation hardening can decrease with dose over certain ranges of fluence. Irradiation induced voids were observed in very low numbers in the Zircaloy-2 materials at the highest fluence.

  3. High dose neutron irradiations of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

    SciTech Connect

    Perez-Bergquist, Alex G.; Nozawa, Takashi; Shih, Chunghao Phillip; Leonard, Keith J.; Snead, Lance Lewis; Katoh, Yutai

    2014-07-01

    Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy is used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300° C. Furthermore, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.

  4. Pixel pitch and particle energy influence on the dark current distribution of neutron irradiated CMOS image sensors.

    PubMed

    Belloir, Jean-Marc; Goiffon, Vincent; Virmontois, Cédric; Raine, Mélanie; Paillet, Philippe; Duhamel, Olivier; Gaillardin, Marc; Molina, Romain; Magnan, Pierre; Gilard, Olivier

    2016-02-22

    The dark current produced by neutron irradiation in CMOS Image Sensors (CIS) is investigated. Several CIS with different photodiode types and pixel pitches are irradiated with various neutron energies and fluences to study the influence of each of these optical detector and irradiation parameters on the dark current distribution. An empirical model is tested on the experimental data and validated on all the irradiated optical imagers. This model is able to describe all the presented dark current distributions with no parameter variation for neutron energies of 14 MeV or higher, regardless of the optical detector and irradiation characteristics. For energies below 1 MeV, it is shown that a single parameter has to be adjusted because of the lower mean damage energy per nuclear interaction. This model and these conclusions can be transposed to any silicon based solid-state optical imagers such as CIS or Charged Coupled Devices (CCD). This work can also be used when designing an optical imager instrument, to anticipate the dark current increase or to choose a mitigation technique. PMID:26907077

  5. High dose neutron irradiations of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

    DOE PAGESBeta

    Perez-Bergquist, Alex G.; Nozawa, Takashi; Shih, Chunghao Phillip; Leonard, Keith J.; Snead, Lance Lewis; Katoh, Yutai

    2014-07-01

    Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy ismore » used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300° C. Furthermore, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.« less

  6. Continued investigation of kinetic aspects of bone mineral metabolism. [determining body calcium by measuring argon after neutron irradiation

    NASA Technical Reports Server (NTRS)

    Palmer, H. E.

    1974-01-01

    The total body calcium in humans was determined by measuring expired Ar-37 after neutron irradiation. The excretion of Ar-37 from humans was found to be much slower than the excretion from rats and dogs, and to be related to the age of a person. A study of the uniformity of the Ar-37 production throughout the thickness of the body was studied using phantoms. The results indicate that it should be possible to obtain a uniformity within plus or minus 3% for the production of Ar-37 per unit of calcium by using a bilateral irradiation. New low background, large volume proportional counters were developed and constructed, for more sensitive measurement of Ar-37 in the expired air from patients. A new irradiation enclosure was developed for measuring total body calcium in rats by the Ar-37 method. With this enclosure the Ar-37 production per gram of calcium is constant with a standard deviation of plus or minus 2.8% for any size rat between 100 and 500 grams. The use of Na-22 as measure of bone replacement in the fractured femur of a dog was not successful.

  7. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT.

    PubMed

    Evans, J F; Blue, T E

    1996-11-01

    Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions "How much?" and "What kind?" of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room , patient "scatterer," and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h-1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel. PMID:8887513

  8. Point defect processes in neutron irradiated Ni, Fe-15Cr-16Ni and Ti-added modified SUS316SS

    NASA Astrophysics Data System (ADS)

    Horiki, M.; Yoshiie, T.; Sato, K.; Xu, Q.

    2013-05-01

    The defect structures in Ni, Fe-15Cr-16Ni and Ti-added modified SUS316SS (modified SUS316) were examined after neutron irradiation below 0.3 dpa by the Japan Materials Testing Reactor and Belgian Reactor 2 to compare their defect structural evolution. The growth behaviour of interstitial-type dislocation loops (I-loops), stacking fault tetrahedra (SFTs) and voids was found to be quite different among these specimens. I-loops developed at lower temperatures in Ni than in Fe-15Cr-16Ni and modified SUS316, and more swelling occurred in Ni than in Fe-15Cr-16Ni. Finally, there were no voids in modified SUS316. These results were analysed in terms of the I-loop energy. A large discrepancy was found between the analytical results and experimental observations for Ni and modified SUS316, which suggests the formation of unfaulted I-loops directly from collision cascades. The growth of SFTs was detected in Fe-15Cr-16Ni and modified SUS316, and can be explained by a change in the dislocation bias of SFTs resulting from the absorption of alloying elements.

  9. 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

  10. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT

    SciTech Connect

    Evans, J.F.; Blue, T.E.

    1996-11-01

    Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions {open_quotes}How much?{close_quotes} and {open_quotes}What kind?{close_quotes} of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room, patient {open_quotes}scatterer,{close_quotes} and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h{sup {minus}1} was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel. 20 refs., 8 figs., 2 tabs.

  11. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

    DOE PAGESBeta

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating themore » irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.« less

  12. High dose neutron irradiation of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

    NASA Astrophysics Data System (ADS)

    Perez-Bergquist, Alejandro G.; Nozawa, Takashi; Shih, Chunghao; Leonard, Keith J.; Snead, Lance L.; Katoh, Yutai

    2015-07-01

    Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy is used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300 °C. Specifically, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.

  13. Fission Neutron Irradiation Effect on Interlaminar Shear Strength of Cyanate Ester Resin Gfrp at RT and 77 K

    NASA Astrophysics Data System (ADS)

    Nishimura, A.; Izumi, Y.; Nishijima, S.; Hemmi, T.; Koizumi, K.; Takeuchi, T.; Shikama, T.

    2010-04-01

    A glass fiber reinforced plastic (GFRP) with cyanate ester resin was fabricated and neutron irradiation tests up to 1×1022 n/m2 of fast neutron with over 0.1 MeV energy were carried out in fission reactor. The fabrication process of cyanate ester GFRP was established and a collaboration network to perform investigations on irradiation effect of superconducting magnet materials was constructed. Three kinds of samples were fabricated. The first was CTD403 GFRP made by NIFS, the second was (cyanate ester+epoxy) GFRP provided by Toshiba, and the last was CTD403 GFRP made by Toshiba. The irradiation was carried out at JRR-3 in Japan Atomic Energy Agency using Rabbit capsules. After the irradiation, short beam tests were conducted at room temperature and 77 K and interlaminar shear strength (ILSS) was evaluated. The irradiation of 1×1021 n/m2 increased ILSS a little but 1×1022 n/m2 irradiation decreased ILSS to around 50 MPa. These tendencies were observed in all three kinds of GFRPs.

  14. Development of A New Class of Fe-3Cr-W(V)Ferritic Steels for Industrial Process Applications

    SciTech Connect

    Sikka, V.J.; Jawad, M.H.

    2005-06-15

    The project, 'Development of a New Class of Fe-Cr-W(V) Ferritic Steels for Industrial Process Applications', was a Cooperative Research and Development Agreement (CRADA) between Oak Ridge National Laboratory (ORNL) and Nooter Corporation. This project dealt with improving the materials performance and fabrication for the hydrotreating reactor vessels, heat recovery systems, and other components for the petroleum and chemical industries. The petroleum and chemical industries use reactor vessels that can approach the ship weights of approximately 300 tons with vessel wall thicknesses of 3 to 8 in. These vessels are typically fabricated from Fe-Cr-Mo steels with chromium ranging from 1.25 to 12% and molybdenum from 1 to 2%. Steels in this composition have great advantages of high thermal conductivity, low thermal expansion, low cost, and properties obtainable by heat treatment. With all of the advantages of Fe-Cr-Mo steels, several issues are faced in design and fabrication of vessels and related components. These issues include the following: (1) low strength properties of current alloys require thicker sections; (2) increased thickness causes heat-treatment issues related to nonuniformity across the thickness and thus not achieving the optimum properties; (3) fracture toughness (ductile-to-brittle transition ) is a critical safety issue for these vessels, and it is affected in thick sections due to nonuniformity of microstructure; (4) PWHT needed after welding and makes fabrication more time-consuming with increased cost; and (5) PWHT needed after welding also limits any modifications of the large vessels in service. The goal of this project was to reduce the weight of large-pressure vessel components (ranging from 100 to 300 tons) by approximately 25% and reduce fabrication cost and improve in-service modification feasibility through development of Fe-3Cr-W(V) steels with combination of nearly a 50% higher strength, a lower DBTT and a higher upper-shelf energy

  15. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    NASA Astrophysics Data System (ADS)

    Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

    2015-12-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

  16. Effect of neutron irradiation on defect evolution in Ti3SiC2 and Ti2AlC

    NASA Astrophysics Data System (ADS)

    Tallman, Darin J.; He, Lingfeng; Garcia-Diaz, Brenda L.; Hoffman, Elizabeth N.; Kohse, Gordon; Sindelar, Robert L.; Barsoum, Michel W.

    2016-01-01

    Herein we report on the characterization of defects formed in polycrystalline Ti3SiC2 and Ti2AlC samples exposed to neutron irradiation - up to 0.1 displacements per atom (dpa) at 350 ± 40 °C or 695 ± 25 °C, and up to 0.4 dpa at 350 ± 40 °C. Black spots are observed in both Ti3SiC2 and Ti2AlC after irradiation to both 0.1 and 0.4 dpa at 350 °C. After irradiation to 0.1 dpa at 695 °C, small basal dislocation loops, with a Burgers vector of b = 1/2 [0001] are observed in both materials. At 9 ± 3 and 10 ± 5 nm, the loop diameters in the Ti3SiC2 and Ti2AlC samples, respectively, were comparable. At 1 × 1023 loops/m3, the dislocation loop density in Ti2AlC was ≈1.5 orders of magnitude greater than in Ti3SiC2, at 3 × 1021 loops/m3. After irradiation at 350 °C, extensive microcracking was observed in Ti2AlC, but not in Ti3SiC2. The room temperature electrical resistivities increased as a function of neutron dose for all samples tested, and appear to saturate in the case of Ti3SiC2. The MAX phases are unequivocally more neutron radiation tolerant than the impurity phases TiC and Al2O3. Based on these results, Ti3SiC2 appears to be a more promising MAX phase candidate for high temperature nuclear applications than Ti2AlC.

  17. Effect of neutron irradiation on defect evolution in Ti3SiC2 and Ti2AlC

    DOE PAGESBeta

    Tallman, Darin J.; He, Lingfeng; Garcia-Diaz, Brenda L.; Hoffman, Elizabeth N.; Kohse, Gordon; Sindelar, Robert L.; Barsoum, Michel W.

    2015-10-23

    Here, we report on the characterization of defects formed in polycrystalline Ti3SiC2 and Ti2AlC samples exposed to neutron irradiation – up to 0.1 displacements per atom (dpa) at 350 ± 40 °C or 695 ± 25 °C, and up to 0.4 dpa at 350 ± 40 °C. Black spots are observed in both Ti3SiC2 and Ti2AlC after irradiation to both 0.1 and 0.4 dpa at 350 °C. After irradiation to 0.1 dpa at 695 °C, small basal dislocation loops, with a Burgers vector of b = 1/2 [0001] are observed in both materials. At 9 ± 3 and 10 ±more » 5 nm, the loop diameters in the Ti3SiC2 and Ti2AlC samples, respectively, were comparable. At 1 × 1023 loops/m3, the dislocation loop density in Ti2AlC was ≈1.5 orders of magnitude greater than in Ti3SiC2, at 3 x 1021 loops/m3. After irradiation at 350 °C, extensive microcracking was observed in Ti2AlC, but not in Ti3SiC2. The room temperature electrical resistivities increased as a function of neutron dose for all samples tested, and appear to saturate in the case of Ti3SiC2. The MAX phases are unequivocally more neutron radiation tolerant than the impurity phases TiC and Al2O3. Based on these results, Ti3SiC2 appears to be a more promising MAX phase candidate for high temperature nuclear applications than Ti2AlC.« less

  18. Determining the shear fracture properties of HIP joints of reduced-activation ferritic/martensitic steel by a torsion test

    NASA Astrophysics Data System (ADS)

    Nozawa, Takashi; Noh, Sanghoon; Tanigawa, Hiroyasu

    2012-08-01

    Hot isostatic pressing (HIP) is a key technology used to fabricate a first wall with cooling channels for the fusion blanket system utilizing a reduced-activation ferritic/martensitic steel. To qualify the HIPped components, small specimen test techniques are beneficial not only to evaluate the thin-wall cooling channels containing the HIP joint but also to use in neutron irradiation studies. This study aims to develop the torsion test method with special emphasis on providing a reasonable and comprehensive method to determine interfacial shear properties of HIP joints during the torsional fracture process. Torsion test results identified that the torsion process shows yield of the base metal followed by non-elastic deformation due to work hardening of the base metal. By considering this work hardening issue, we propose a reasonable and realistic solution to determine the torsional yield shear stress and the ultimate torsional shear strength of the HIPped interface. Finally, a representative torsion fracture process was identified.

  19. The impact of different flooding periods on the dynamics of pore water concentrations of As, Cr, Mo and V in a contaminated floodplain soil - results of a lysimeter study

    NASA Astrophysics Data System (ADS)

    Rupp, Holger; Meissner, Ralph; Shaheen, Sabry; Rinklebe, Jörg

    2014-05-01

    Trace elements and arsenic (As) were transported with water during inundation in floodplain ecosystems, where they settled down and accumulated predominantly in depressions and low-lying terraces. Highly variable hydrological conditions in floodplains can affect the dynamics of pollutants. The impact of different flooding/drying periods on the temporal dynamics of pore water concentrations of As, Cr, Mo and V as a function of soil EH/pH changes and dynamics of DOC, Fe, Mn and SO42- was studied in a contaminated floodplain soil collected at the Elbe River (Germany). A specific groundwater lysimeter technique with two separate small lysimeter vessels served as replicates was used for this study. The groundwater level inside the lysimeters was controlled to simulate long term and short term flooding/drying. The long term (LT) flooding scenario consists of 94 days of flooding followed by similar drying term. The short term (ST) flooding/drying scenario comprises 21 days and was six times repeated. The entire experimental period (LT_ST) was about 450 days. Flooding of the soil caused a significant decrease of EH and pH. Concentrations of soluble As, Cr, Fe, Mn, Mo and DOC were higher under reducing conditions than under oxidizing conditions in LT. However, As and Cr tended to be mobilized under oxidizing conditions during ST, which might be due to slow kinetics of the redox reaction of As and Cr. Dynamics of Mo were more affected by changes of EH/pH as compared to As, Cr and V and governed mainly by Fe-Mn chemistry. Concentrations of V in ST were higher than in LT and were controlled particularly by pH and chemistry of Fe. The interactions between the elements and carriers studied were stronger during long flood-dry-cycles than during short cycles, which confirmed our hypothesis. We conclude that the dynamics of As, Cr, Mo and V are determined by the length of time soils are exposed to flooding, because drivers of element mobility need a certain time to provoke

  20. The development of ferritic-martensitic steels with reduced long-term activation

    NASA Astrophysics Data System (ADS)

    Ehrlich, K.; Kelzenberg, S.; Röhrig, H.-D.; Schäfer, L.; Schirra, M.

    1994-09-01

    Ferritic-martensitic 9-12% CrMoVNb steels of MANET type possess a number of advantageous properties for fusion reactor application. Their optimization has led to improved creep and fracture-toughness properties. New 9-10% CrWVTa alloys have been developed by KfK/IMF in collaboration with the SAARSTAHL GmbH which have a reduced long-term activation and show in addition superior fracture toughness properties. The calculation of dose rate and other radiological parameters with the presently available FISPACT/EAF codes, extended by KfK files for sequential reactions has shown that the long-term dose-rate in these alloys is governed by the remaining 'impurity level' of Nb and the alloying elements W and Ta. Sequential reactions — though relevant for single alloying elements like Cr, Mn, V and N — provide only a second order effect in Fe-based alloys. A challenge for the future materials development is the production of alloys with the desired narrow specification of elements and impurities, which necessitates new ways of steelmaking.

  1. 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.

  2. 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.

  3. Magnetic and electrical properties of flux grown single crystals of Ln{sub 6}M{sub 4}Al{sub 43} (Ln=Gd, Yb; M=Cr, Mo, W)

    SciTech Connect

    Kangas, Michael J.; Treadwell, LaRico J.; Haldolaarachchige, Neel; McAlpin, Jacob D.; Young, David P.; Chan, Julia Y.

    2013-01-15

    Millimeter-sized single crystals of Ln{sub 6}M{sub 4}Al{sub 43} (Ln=Gd, Yb; M=Cr, Mo, W) were successfully grown with a molten aluminum flux. Synthetic conditions and physical properties for single crystals of all six analogs are discussed. The compounds exhibit metallic resistivity with room temperature values between 0.1 and 0.6 m{Omega}-cm. The Yb analogs are Pauli paramagnets with the Yb ion adopting the nonmagnetic divalent configuration (Yb{sup 2+}). Gd{sub 6}Cr{sub 4}Al{sub 43}, Gd{sub 6}Mo{sub 4}Al{sub 43}, and Gd{sub 6}W{sub 4}Al{sub 43} appear to order antiferromagnetically at 19, 15, and 15 K, respectively. - Graphical abstract: The crystal structure of Yb{sub 6}Cr{sub 4}Al{sub 43}. The light and dark green polyhedra show the chromium sublattice. Highlights: Black-Right-Pointing-Pointer Single crystals up to 0.5 cm in length were grown with a molten aluminum flux. Black-Right-Pointing-Pointer Physical property measurements were conducted on single crystals. Black-Right-Pointing-Pointer Gadolinium analogs appear to order antiferromagnetically with positive {theta}. Black-Right-Pointing-Pointer All analogs show metallic resistivity.

  4. Influence of boron vacancies on phase stability, bonding and structure of MB 2 (M  =  Ti, Zr, Hf, V, Nb, Ta, Cr, Mo, W) with AlB2 type structure

    NASA Astrophysics Data System (ADS)

    Dahlqvist, Martin; Jansson, Ulf; Rosen, Johanna

    2015-11-01

    Transition metal diborides in hexagonal AlB2 type structure typically form stable MB 2 phases for group IV elements (M  =  Ti, Zr, Hf). For group V (M  =  V, Nb, Ta) and group VI (M  =  Cr, Mo, W) the stability is reduced and an alternative hexagonal rhombohedral MB 2 structure becomes more stable. In this work we investigate the effect of vacancies on the B-site in hexagonal MB 2 and its influence on the phase stability and the structure for TiB2, ZrB2, HfB2, VB2, NbB2, TaB2, CrB2, MoB2, and WB2 using first-principles calculations. Selected phases are also analyzed with respect to electronic and bonding properties. We identify trends showing that MB 2 with M from group V and IV are stabilized when introducing B-vacancies, consistent with a decrease in the number of states at the Fermi level and by strengthening of the B-M interaction. The stabilization upon vacancy formation also increases when going from M in period 4 to period 6. For TiB2, ZrB2, and HfB2, introduction of B-vacancies have a destabilizing effect due to occupation of B-B antibonding orbitals close to the Fermi level and an increase in states at the Fermi level.

  5. Effect of scanning speeds on microstructure and wear behavior of laser-processed NiCr-Cr3C2-MoS2-CeO2 on 38CrMoAl steel

    NASA Astrophysics Data System (ADS)

    Sun, Guifang; Tong, Zhaopeng; Fang, Xiaoyu; Liu, Xiaojun; Ni, Zhonghua; Zhang, Wei

    2016-03-01

    Self-lubricating wear-resistant NiCr-Cr3C2-MoS2-CeO2 layers were fabricated on 38CrMoAl extruder screws by laser processing. The effect of scanning speeds on microstructure, phases, microhardness, and wear behavior was investigated. The obtained results indicate that the laser-processed layers had fine and nonuniform microstructures with undissolved MoS2 particles distributed on the matrix. With an increase of the laser-scanning speeds, the microstructures changed from hypoeutectic to hypereutectic, volume fraction of martensite increased, microhardness increased, and thickness and friction coefficients of the layers decreased. Wear resistance of the optimized layer was increased by 29.76 times compared with that of the substrate. The undissolved MoS2 was separated from the matrix on loading. In addition to the grain-refining and solution-strengthening effects, oxide films formed on the surface of the layers shielded them and enhanced their wear resistance. The crack or fracture behavior of the laser-processed layers on loading was determined by its toughness, which also had an important effect on the wear behavior of the processed layers.

  6. Influence of boron vacancies on phase stability, bonding and structure of MB₂ (M  =  Ti, Zr, Hf, V, Nb, Ta, Cr, Mo, W) with AlB₂ type structure.

    PubMed

    Dahlqvist, Martin; Jansson, Ulf; Rosen, Johanna

    2015-11-01

    Transition metal diborides in hexagonal AlB2 type structure typically form stable MB2 phases for group IV elements (M  =  Ti, Zr, Hf). For group V (M  =  V, Nb, Ta) and group VI (M  =  Cr, Mo, W) the stability is reduced and an alternative hexagonal rhombohedral MB2 structure becomes more stable. In this work we investigate the effect of vacancies on the B-site in hexagonal MB2 and its influence on the phase stability and the structure for TiB2, ZrB2, HfB2, VB2, NbB2, TaB2, CrB2, MoB2, and WB2 using first-principles calculations. Selected phases are also analyzed with respect to electronic and bonding properties. We identify trends showing that MB2 with M from group V and IV are stabilized when introducing B-vacancies, consistent with a decrease in the number of states at the Fermi level and by strengthening of the B-M interaction. The stabilization upon vacancy formation also increases when going from M in period 4 to period 6. For TiB2, ZrB2, and HfB2, introduction of B-vacancies have a destabilizing effect due to occupation of B-B antibonding orbitals close to the Fermi level and an increase in states at the Fermi level. PMID:26445165

  7. Elaboration, étude structurale et analyse CHARDI et BVS d’une nouvelle variété β-Na9Cr(MoO4)6 de type alluaudite

    PubMed Central

    Sonni, Manel; Marzouki, Riadh; Zid, Mohamed Faouzi; Souilem, Amira

    2016-01-01

    The title compound, nonasodium chromium(III) hexakis[molybdate(VI)], β-Na9CrMo6O24, was prepared by solid-state reactions. This alluaudite-type structure is constituted of infinite layers formed by links between M 2O10 (M = C/Na) dimers and MoO4 tetra­hedra. The Na+ and Cr3+ cations are located in the same site with, respectively, 0.25 and 0.75 occupancies. The layers are connected to each other through MoO4 sharing corners, resulting an in open three-dimensional framework with hexa­gonal-form cavities occupied by Na+ cations. The proposed structural model is supported by charge-distribution (CHARDI) and bond-valence-sum (BVS) analysis. All atoms are on general positions except for one Mo, two Na (site symmetry 2) and another Na site (site symmetry -1). A comparison is made with the similar structures Na4Co(MoO4)3, Na2Ni(MoO4)2, Cu1.35Fe3(PO4)3 and NaAgFeMn2(PO4)3. PMID:27308053

  8. Effects of fabrication practices and techniques on the corrosion and mechanical properties of Ni-Cr-Mo nickel based alloys UNS N10276, N06022, N06686, and N06625

    SciTech Connect

    Hinshaw, E.B.; Crum, J.R.

    1996-11-01

    Ni-Cr-Mo alloys have excellent resistance to both oxidizing and reducing type environments; however, heat treating, surface condition, welding, and type of welding consumable can have a significant affect on the corrosion resistance and mechanical properties of these alloys. It is also important when performing standard ASTM intergranular corrosion tests on welded test coupons to make an accurate comparison of alloys being tested. A standard weld procedure and consistent post-weld sample conditioning method should be incorporated into the comparison test program. An evaluation of the effect of various fabrication practices on the corrosion resistance of the alloy was performed using accelerated corrosion tests ASTM G28B. The fabrication conditions examined were as-welded, welded-pickled, welded-annealed-pickled, welded annealed ground, welded-ground, using over matching filler metals, and various levels of heat input. In addition to fabrication techniques, the effect of ASTM G28B test duration on corrosion rates of UNS N10276, N06022, N06686, and N06625 was evaluated. ASTM G28A intergranular corrosion and mechanical testing using welded coupons of UNS N06625 were also performed to determine the affect of post-weld annealing and aging heat treatments on the corrosion resistance and mechanical properties of UNS N06625.

  9. Alloy Design and Development of Cast Cr-W-V Ferritic Steels for Improved High-Temperature Strength for Power Generation Applications

    SciTech Connect

    Klueh, R L; Maziasz, P J; Vitek, J M; Evans, N D; Hashimoto, N

    2006-09-23

    Economic and environmental concerns demand that the power-generation industry seek increased efficiency for gas turbines. Higher efficiency requires higher operating temperatures, with the objective temperature for the hottest sections of new systems {approx} 593 C, and increasing to {approx} 650 C. Because of their good thermal properties, Cr-Mo-V cast ferritic steels are currently used for components such as rotors, casings, pipes, etc., but new steels are required for the new operating conditions. The Oak Ridge National Laboratory (ORNL) has developed new wrought Cr-W-V steels with 3-9% Cr, 2-3% W, 0.25% V (compositions are in wt.%), and minor amounts of additional elements. These steels have the strength and toughness required for turbine applications. Since cast alloys are expected to behave differently from wrought material, work was pursued to develop new cast steels based on the ORNL wrought compositions. Nine casting test blocks with 3, 9, and 11% Cr were obtained. Eight were Cr-W-V-Ta-type steels based on the ORNL wrought steels; the ninth was COST CB2, a 9Cr-Mo-Co-V-Nb cast steel, which was the most promising cast steel developed in a European alloy-development program. The COST CB2 was used as a control to which the new compositions were compared, and this also provided a comparison between Cr-W-V-Ta and Cr-Mo-V-Nb compositions. Heat treatment studies were carried out on the nine castings to determine normalizing-and-tempering treatments. Microstructures were characterized by both optical and transmission electron microscopy (TEM). Tensile, impact, and creep tests were conducted. Test results on the first nine cast steel compositions indicated that properties of the 9Cr-Mo-Co-V-Nb composition of COST CB2 were better than those of the 3Cr-, 9Cr-, and 11Cr-W-V-Ta steels. Analysis of the results of this first iteration using computational thermodynamics raised the question of the effectiveness in cast steels of the Cr-W-V-Ta combination versus the Cr-Mo

  10. Effect of neutron irradiation on the microstructure and the mechanical and corrosion properties of the ultrafine-grained stainless Cr-Ni steel

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.; Gusev, M. N.; Tsai, K. V.; Yarovchuk, A. V.; Rybalchenko, O. V.; Enikeev, N. A.; Valiev, R. Z.; Dobatkin, S. V.

    2015-12-01

    It has been revealed that the neutron irradiation of ultrafine-grained (UFG) 08Kh18N10T steel after severe plastic deformation (SPD) does not lead to the appearance of defects of radiation origin up to a fluence of 2 × 1020 n/cm2 (~0.05 dpa) and that the strength properties of the material are retained after irradiation. At the same time, this irradiation reduces the corrosion resistance of the steel in a chlorine-containing medium, especially after heating at 550°C with a holding for 1 h after SPD.

  11. Correlated evolution of barrier capacitance charging, generation, and drift currents and of carrier lifetime in Si structures during 25 MeV neutrons irradiation

    SciTech Connect

    Gaubas, E.; Ceponis, T.; Jasiunas, A.; Uleckas, A.; Vaitkus, J.; Cortina, E.; Militaru, O.

    2012-12-03

    The in situ examination of barrier capacitance charging, of generation and drift currents, and of carrier lifetime in Si structures during 25 MeV neutrons irradiation has been implemented to correlate radiation induced changes in carrier recombination, thermal release, and drift characteristics and to clarify their impact on detector performance. It has been shown that microwave probed photo-conductivity technique implemented in contact-less and distant manner can be a powerful tool for examination in wide dynamic range of carrier lifetime modified by radiation defects and for rather precise prediction of detector performance.

  12. 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

  13. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    SciTech Connect

    Lillo, Thomas; Rooyen, Isabella Van

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all

  14. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory's AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ∼23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ∼24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (∼10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not

  15. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Hashimoto, N.; Sokolov, M. A.; Maziasz, P. J.; Shiba, K.; Jitsukawa, S.

    2006-10-01

    In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10-12 dpa at 300 and 400 °C and in the Fast Flux Test Facility (FFTF) to 15 dpa at 393 °C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed.

  16. The effects of neutron irradiation on shear properties of monolayered PyC and multilayered PyC/SiC interfaces of SiC/SiC composites

    NASA Astrophysics Data System (ADS)

    Nozawa, T.; Katoh, Y.; Snead, L. L.

    2007-08-01

    The effect of neutron irradiation on mechanical properties at the fiber/matrix interface of SiC/SiC composites was evaluated. The materials investigated were Hi-Nicalon™ Type-S fiber reinforced chemically vapor infiltrated SiC matrix composites with varied interphases: monolayered pyrolytic carbon (PyC) or multilayered PyC/SiC. The neutron fluence was 7.7 × 10 25 n/m 2 ( E > 0.1 MeV), and the irradiation temperature was 800 °C. Interfacial shear properties were evaluated by the fiber push-out test method. A modified shear-lag model was applied to analyze the interfacial shear parameters. Test results indicate that the interfacial debond shear strength and the interfacial friction stress for the multilayer composites were significantly degraded by irradiation. Nevertheless, the multilayer composites retained sufficient interfacial shear properties so that overall composite strength after neutron irradiation was unaffected. The actual mechanism of interphase property decrease for the multilayer composites is unknown. The interfacial shear properties of the irradiated monolayer composites appear unaffected.

  17. The medical-irradiation characteristics for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    At the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor, the mix irradiation of thermal and epi-thermal neutrons, and the solo irradiation of epi-thermal neutrons are available additionally to the thermal neutron irradiation, and then the neutron capture therapy (NCT) at this facility became more flexible, after the update in 1996. The estimation of the depth dose distributions in NCT clinical irradiation, were performed for the standard irradiation modes of thermal, mixed and epi-thermal neutrons, from the both sides of experiment and calculation. On the assumption that the 10B concentration in tumor part was 40 ppm and the ratio of tumor to normal tissue was 3.5, the advantage depth were estimated to 5.4, 6.0, and 8.0, for the respective standard irradiation modes. It was confirmed that the various irradiation conditions can be selected according to the target-volume conditions, such as size, depth, etc. Besides, in the viewpoint of the radiation shielding for patient, it was confirmed that the whole-body exposure is effectively reduced by the new clinical collimators, compared with the old one. PMID:12408307

  18. Positron Annihilation Lifetime Spectroscopy Study of Neutron Irradiated High Temperature Superconductors YBa2Cu3O7-δ for Application in Fusion Facilities

    NASA Astrophysics Data System (ADS)

    Veterníková, J.; Chudý, M.; Slugeň, V.; Eisterer, M.; Weber, H. W.; Sojak, S.; Petriska, M.; Hinca, R.; Degmová, J.; Sabelová, V.

    2012-02-01

    This study focuses on the crystallographic defects introduced by neutron irradiation and the resulting changes of the superconducting properties in the high temperature superconductor YBa2Cu3O7-δ. This material is considered to be most promising for magnet systems in future fusion reactors. Two different bulk samples, pure non-doped YBa2Cu3O7-δ (YBCO) and multi-seed YBa2Cu3O7-δ doped by platinum (MS2F) were studied prior to and after irradiation in the TRIGA MARK II reactor in Vienna. Neutron irradiation is responsible for a significant enhancement of the critical current densities as well as for a reduction in critical temperature. The accumulation of small open volume defects (<0.5 nm) partially causes those changes. These defects were studied by positron annihilation lifetime spectroscopy at room temperature. A high concentration of Cu-O di-vacancies was found in both samples, which increased with neutron fluence. The defect concentration was significantly reduced after a heat treatment.

  19. Study of the structure and properties of metal of the major steam lines of a CCGT-420 unit made from high-chromium X10CrMoVNb9-1 (P91) steel

    NASA Astrophysics Data System (ADS)

    Grin', E. A.; Anokhov, A. E.; Pchelintsev, A. V.; Krüger, E.-T.

    2016-07-01

    The technology of manufacture of live steam lines and hot reheat lines at FINOW Rohrsysteme GmbH are discussed. These pipelines are designed for high-performance CCGT units and are made from high-chromium martensitic steel X10CrMoVNb9-1 (P91). The principles of certification and evaluation of conformance of thermal and mechanical equipment made from new construction materials with the TRCU 032-2013 technical regulation of the Customs Union are detailed. The requirements outlined in Russian and international regulatory documents regarding the manufacture of pipes and semifinished products for pipeline systems are compared. The characteristic features of high-chromium martensitic steel, which define the requirements for its heat treatment and welding, are outlined. The methodology and the results of a comprehensive analysis of metal of pipes, fittings, and weld joints of steam lines are presented. It is demonstrated that the short-term mechanical properties of metal (P91 steel) of pipes, bends, and weld joints meet the requirements of European standards and Russian technical specifications. The experimental data on long-term strength of metal of pipes from a live steam line virtually match the corresponding reference curve from the European standard, while certain experimental points for metal of bends of this steam line and metal of pipes and bends from a hot reheat line lie below the reference curve, but they definitely stay within the qualifying (20%) interval of the scatter band. The presence of a weakened layer in the heat-affected zone of weld joints of steel P91 is established. It is shown that the properties of this zone govern the short-term and long-term strength of weld joints in general. The results of synthesis and analysis of research data support the notion that the certification testing of steam lines and other equipment made from chromium steels should necessarily involve the determination of long-term strength parameters.

  20. Retained Austenite Decomposition and Carbide Formation During Tempering a Hot-Work Tool Steel X38CrMoV5-1 Studied by Dilatometry and Atom Probe Tomography

    NASA Astrophysics Data System (ADS)

    Lerchbacher, Christoph; Zinner, Silvia; Leitner, Harald

    2012-12-01

    The microstructural development of a hot-work tool steel X38CrMoV5-1 during continuous heating to tempering temperature has been investigated with the focus on the decomposition of retained austenite (Stage II) and carbide formation (Stages III and IV). Investigations have been carried out after heating to 673.15 K, 773.15 K, 883.15 K (400 °C, 500 °C, 610 °C) and after a dwell time of 600 seconds at 883.15 K (610 °C). Dilatometry and atom probe tomography were used to identify tempering reactions. A distinctive reaction takes place between 723.15 K and 823.15 K (450 °C and 550 °C) which is determined to be the formation of M3C from transition carbides. Stage II could be evidenced with the atom probe results and indirectly with dilatometry, indicating the formation of new martensite during cooling. Retained austenite decomposition starts with the precipitation of alloy carbides formed from nanometric interlath retained austenite films which are laminary arranged and cause a reduction of the carbon content within the retained austenite. Preceding enrichment of substitutes at the matrix/carbide interface in the early stages of Cr7C3 alloy carbide formation could be visualised on the basis of coarse M3C carbides within the matrix. Atom probe tomography has been found to be very useful to complement dilatational experiments in order to characterise and identify microstructural changes.

  1. High strength ferritic alloy

    DOEpatents

    Hagel, William C.; Smidt, Frederick A.; Korenko, Michael K.

    1977-01-01

    A high-strength ferritic alloy useful for fast reactor duct and cladding applications where an iron base contains from about 9% to about 13% by weight chromium, from about 4% to about 8% by weight molybdenum, from about 0.2% to about 0.8% by weight niobium, from about 0.1% to about 0.3% by weight vanadium, from about 0.2% to about 0.8% by weight silicon, from about 0.2% to about 0.8% by weight manganese, a maximum of about 0.05% by weight nitrogen, a maximum of about 0.02% by weight sulfur, a maximum of about 0.02% by weight phosphorous, and from about 0.04% to about 0.12% by weight carbon.

  2. Excimer laser ablation of ferrites

    NASA Astrophysics Data System (ADS)

    Tam, A. C.; Leung, W. P.; Krajnovich, D.

    1991-02-01

    Laser etching of ferrites was previously done by scanning a focused continuous-wave laser beam on a ferrite sample in a chemical environment. We study the phenomenon of photo-ablation of Ni-Zn or Mn-Zn ferrites by pulsed 248-nm KrF excimer laser irradiation. A transfer lens system is used to project a grating pattern of a mask irradiated by the pulsed KrF laser onto the ferrite sample. The threshold fluence for ablation at the ferrite surface is about 0.3 J/cm2. A typical fluence of 1 J/cm2 is used. The etched grooves produced are typically 20-50 μm wide, with depths achieved as deep as 70 μm . Groove straightness is good as long as a sharp image is projected onto the sample surface. The wall angle is steeper than 60 degrees. Scanning electron microscopy of the etched area shows a ``glassy'' skin with extensive microcracks and solidified droplets being ejected that is frozen in action. We found that this skin can be entirely removed by ultrasonic cleaning. A fairly efficient etching rate of about 10 nm/pulse for a patterned area of about 2 mm×2 mm is obtained at a fluence of 1 J/cm2. This study shows that projection excimer laser ablation is useful for micromachining of ferrite ceramics, and indicates that a hydrodynamic sputtering mechanism involving droplet emission is a cause of material removal.

  3. Why neutron guides may end up breaking down? Some results on the macroscopic behaviour of alkali-borosilicate glass support plates under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Boffy, R.; Kreuz, M.; Beaucour, J.; Köster, U.; Bermejo, F. J.

    2015-09-01

    In this paper we report on a first part of a study on the mechanisms leading to brittle fracture in neutron guides made of glass as structural element. Such devices are widely used to deliver thermal and cold neutron beams to experimental lines in most large neutron research facilities. We present results on macroscopic properties of samples of guide glass substrates which are subjected to neutron irradiation at relatively large fluences. The results show a striking dependence of some of the macroscopic properties such as density, shape or surface curvature upon the specific chemical composition of a given glass. The relevance of the present findings for the installation of either replacement guides at the existing facilities or for the deployment of instruments for ongoing projects such as the European Spallation Source is briefly discussed.

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

    SciTech Connect

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

    2012-01-01

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

  5. 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.

  6. Effect of neutron irradiation on the London penetration depth for polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor

    NASA Technical Reports Server (NTRS)

    Ossandon, J. G.; Thompson, J. R.; Sun, Yang Ren; Christen, D. K.; Chakoumakos, B. C.

    1995-01-01

    Magnetization studies of polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor, prior to and after neutron irradiation, showed an increase in J(sub c) due to irradiation damage. Analysis of the equilibrium magnetization revealed significant increases in other more fundamental properties. In particular, the London penetration depth increased by approximately 15 percent following irradiation with 8 x 10(exp 16) neutrons/sq cm. Corresponding changes were observed in the upper critical magnetic field H(sub c2). However, the most fundamental thermodynamic property, the superconductive condensation energy F(sub c), was unaffected by the moderate level of neutron-induced damage.

  7. Thermophysical and mechanical properties of near-stoichiometric fiber CVI SiC/SiC composites after neutron irradiation at elevated temperatures

    NASA Astrophysics Data System (ADS)

    Katoh, Yutai; Snead, Lance L.; Nozawa, Takashi; Kondo, Sosuke; Busby, Jeremy T.

    2010-08-01

    Thermophysical and mechanical properties of high purity chemically vapor-deposited (CVD) SiC and chemically vapor-infiltrated SiC matrix, pyrocarbon/SiC multilayered interphase composites with Hi-Nicalon™ Type-S and Tyranno™-SA3 SiC fibers were evaluated following neutron irradiation. Specimens including statistically significant population of tensile bars were irradiated up to 5.3 displacement-per-atom at ˜220 to ˜1080 °C in the Advanced Test Reactor at Idaho National Laboratory and High Flux Isotope Reactor at Oak Ridge National Laboratory. Thermal diffusivity/conductivity of all materials decreased during irradiation. The reciprocal thermal diffusivity linearly increased with temperature from ambient to the irradiation temperature. The magnitude of defect thermal resistance was distinctively different among materials and its ranking was Hi-Nicalon™ Type-S > Tyranno™-SA3 > CVD SiC regardless of irradiation condition. Dynamic Young's modulus decrease for the irradiated CVD SiC exhibited explicit correlation with swelling. No significant effects of neutron irradiation on tensile properties of the composites were revealed, except for an anomaly case for the Hi-Nicalon™ Type-S composite irradiated in a specific condition. According to the single filament tensile evaluation, fibers of both types retained the original strength during irradiation at intermediate temperatures but significantly deteriorated during bare fiber irradiation at ˜910 °C. However, fiber strength deterioration was not observed when irradiated in composite form. Irradiation effects on the fiber-matrix interface properties were discussed based on results from the composite and single filament tensile tests, the hysteresis analysis, and the fracture surface examination.

  8. Quadruple metal-metal bonds with strong donor ligands. Ultraviolet photoelectron spectroscopy of M{sub 2}(form){sub 4} (M = Cr, Mo, W; form = N,N{prime}-diphenylformamidinate)

    SciTech Connect

    Lichtenberger, D.L.; Lynn, M.A.; Chisholm, M.H.

    1999-12-29

    The He I photoelectron spectra of M{sub 2}(form){sub 4}(M = Cr, Mo, W; form - N,N{prime}-diphenylformamidinate) and Mo{sub 2}(cyform){sub 4} (cyform = N,N{prime}-dicyclohexylformamidinate) are presented. For comparison, the Ne I, He I, and He II photoelectron spectra of Mo{sub 2}(p-CH{sub 3}-form){sub 4} have also been obtained. The valence ionization features of these molecules are interpreted based on (1) the changes that occur with the metal and ligand substitutions, (2) the changes in photoelectron cross sections with excitation source, and (3) the changes from previously studied dimetal complexes. These photoelectron spectra are useful for revealing the effects that better electron donor ligands have on the valence electronic structure of M{sub 2}(L-L){sub 4} systems. Comparison with the He I spectra of the isoelectronic M{sub 2}(O{sub 2}CCH{sub 3}){sub 4} compounds is particularly revealing. Unlike with the more electron-withdrawing acetate ligand, several formamidinate-based ionizations derived from the nitrogen p{sub {pi}} orbitals occur among the metal-metal {sigma}, {pi}, and {delta} ionization bands. Although these formamidinate-based levels are close in energy to the occupied metal-metal bonds, they have little direct mixing interaction with them. The shift of the metal-metal bond ionizations to lower ionization energies for the formamidinate systems is primarily a consequence of the lower electronegativity of the ligand and the better {pi} donation into empty metal levels. The metal-metal {delta} orbital experiences some additional net bonding interaction with ligand orbitals of the same symmetry. Also, an additional bonding interaction from ligand-to-metal electron donation to the {delta}* orbital is identified. These spectra suggest a greater degree of metal-ligand covalency than in the related M{sub 2}(O{sub 2}CCH{sub 3}){sub 4} systems. Fenske-Hall molecular orbital and density functional (ADF) calculations agree with the assignment and

  9. Using Solution- and Solid-State S K-edge X-ray Absorption Spectroscopy with Density Functional Theory to Evaluate M–S Bonding for MS42- (M = Cr, Mo, W) Dianions

    PubMed Central

    Olson, Angela C.; Keith, Jason M.; Batista, Enrique R.; Boland, Kevin S.; Daly, Scott R.; Kozimor, Stosh A.; MacInnes, Molly M.; Martin, Richard L.; Scott, Brian L.

    2014-01-01

    Herein, we have evaluated relative changes in M–S electronic structure and orbital mixing in Group 6 MS42- dianions using solid- and solution-phase S K-edge X-ray absorption spectroscopy (XAS; M = Mo, W), as well as density functional theory (DFT; M = Cr, Mo, W) and time-dependent density functional theory (TDDFT) calculations. To facilitate comparison with solution measurements (conducted in acetonitrile), theoretical models included gas-phase calculations as well as those that incorporated an acetonitrile dielectric, the latter of which provided better agreement with experiment. Two pre-edge features arising from S 1s → e* and t2* electron excitations were observed in the S K-edge XAS spectra and were reasonably assigned as 1A1 → 1T2 transitions. For MoS42-, both solution-phase pre-edge peak intensities were consistent with results from the solid-state spectra. For WS42-, solution- and solid-state pre-edge peak intensities for transitions involving e* were equivalent, while transitions involving the t2* orbitals were less intense in solution. Experimental and computational results have been presented in comparison to recent analyses of MO42- dianions, which allowed M–S and M–O orbital mixing to be evaluated as the principle quantum number (n) for the metal valence d orbitals increased (3d, 4d, 5d). Overall, the M–E (E = O, S) analyses revealed distinct trends in orbital mixing. For example, as the Group 6 triad was descended, e* (π*) orbital mixing remained constant in the M–S bonds, but increased appreciably for M–O interactions. For the t2* orbitals (σ* + π*), mixing decreased slightly for M–S bonding and increased only slightly for the M–O interactions. These results suggested that the metal and ligand valence orbital energies and radial extensions delicately influenced the orbital compositions for isoelectronic ME42- (E = O, S) dianions. PMID:25311904

  10. Using solution- and solid-state S K-edge X-ray absorption spectroscopy with density functional theory to evaluate M-S bonding for MS4(2-) (M = Cr, Mo, W) dianions.

    PubMed

    Olson, Angela C; Keith, Jason M; Batista, Enrique R; Boland, Kevin S; Daly, Scott R; Kozimor, Stosh A; MacInnes, Molly M; Martin, Richard L; Scott, Brian L

    2014-12-14

    Herein, we have evaluated relative changes in M-S electronic structure and orbital mixing in Group 6 MS4(2-) dianions using solid- and solution-phase S K-edge X-ray absorption spectroscopy (XAS; M = Mo, W), as well as density functional theory (DFT; M = Cr, Mo, W) and time-dependent density functional theory (TDDFT) calculations. To facilitate comparison with solution measurements (conducted in acetonitrile), theoretical models included gas-phase calculations as well as those that incorporated an acetonitrile dielectric, the latter of which provided better agreement with experiment. Two pre-edge features arising from S 1s → e* and t electron excitations were observed in the S K-edge XAS spectra and were reasonably assigned as (1)A1 → (1)T2 transitions. For MoS4(2-), both solution-phase pre-edge peak intensities were consistent with results from the solid-state spectra. For WS4(2-), solution- and solid-state pre-edge peak intensities for transitions involving e* were equivalent, while transitions involving the t orbitals were less intense in solution. Experimental and computational results have been presented in comparison to recent analyses of MO4(2-) dianions, which allowed M-S and M-O orbital mixing to be evaluated as the principle quantum number (n) for the metal valence d orbitals increased (3d, 4d, 5d). Overall, the M-E (E = O, S) analyses revealed distinct trends in orbital mixing. For example, as the Group 6 triad was descended, e* (π*) orbital mixing remained constant in the M-S bonds, but increased appreciably for M-O interactions. For the t orbitals (σ* + π*), mixing decreased slightly for M-S bonding and increased only slightly for the M-O interactions. These results suggested that the metal and ligand valence orbital energies and radial extensions delicately influenced the orbital compositions for isoelectronic ME4(2-) (E = O, S) dianions. PMID:25311904

  11. Ferrite attenuator modulation improves antenna performance

    NASA Technical Reports Server (NTRS)

    Hooks, J. C.; Larson, S. G.; Shorkley, F. H.; Williams, B. T.

    1970-01-01

    Ferrite attenuator inserted into appropriate waveguide reduces the gain of the antenna element which is causing interference. Modulating the ferrite attenuator to change the antenna gain at the receive frequency permits ground tracking until the antenna is no longer needed.

  12. Effect of ferrite on cast stainless steels

    SciTech Connect

    Nadezhdin, A.; Cooper, K. ); Timbers, G. . Kraft Pulp Division)

    1994-09-01

    Premature failure of stainless steel castings in bleach washing service is attributed to poor casting quality high porosity and to a high ferrite content, which makes the castings susceptible to corrosion by hot acid chloride solutions. A survey of the chemical compositions and ferrite contents of corrosion-resistant castings in bleach plants at three pulp mills found high [delta]-ferrite levels in the austenitic matrix due to the improper balance between austenite and ferrite stabilizers.

  13. Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

    2013-01-01

    Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3 T M ( T M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

  14. Safe Use Limits for Advanced Ferritic Steels in Ultra-Supercritical Power Boilers.

    SciTech Connect

    Swindeman, RW

    2003-11-03

    In 2000, a Cooperative Research and Development Agreement (CRADA) was undertaken between the Oak Ridge National Laboratory (ORNL) and the Babcock & Wilcox Company to examine the databases for advanced ferritic steels and determine the safe limits for operation in supercritical steam power boilers. The materials of interest included the vanadium-modified 9-12% Cr steels with 1-2% Mo or W. The first task involved a review of pertinent information and the down-selection of a steel of special interest. The long-time database for 9Cr-1Mo-V steel was found to be most satisfactory for the examinations, and this steel was taken to be representative of the group. The second task involved the collection of aged and service exposed samples for metallurgical and mechanical testing. Here, aged samples to 75,000 hours, laboratory-tested samples to 83,000 hours, and service-exposed sample with up to 143,000 hours exposure were collected. The third task involved mechanical testing of exposed samples. Creep-rupture testing to long times was undertaken. Variable stress and temperature testing was included. Results were compared against the prediction of damage models. These models seemed to be adequate for life prediction. The fourth task involved the metallurgical examination of exposed specimens. Changes in microstructure were compared against published information on the evolution of microstructures in 9Cr-Mo-V steels and the results were found to be consistent with expectations. The fifth task involved a survey of steam and fireside corrosion. Data from the service-exposed tubing was examined, and a literature survey was undertaken as part of an activity in support of ultra-supercritical steam boiler technology. The corrosion study indicated some concerns about long-time fireside corrosion and suggested temperature limits were needed for corrosive coal ash conditions.

  15. Articles comprising ferritic stainless steels

    DOEpatents

    Rakowski, James M.

    2016-06-28

    An article of manufacture comprises a ferritic stainless steel that includes a near-surface region depleted of silicon relative to a remainder of the ferritic stainless steel. The article has a reduced tendency to form an electrically resistive silica layer including silicon derived from the steel when the article is subjected to high temperature oxidizing conditions. The ferritic stainless steel is selected from the group comprising AISI Type 430 stainless steel, AISI Type 439 stainless steel, AISI Type 441 stainless steel, AISI Type 444 stainless steel, and E-BRITE.RTM. alloy, also known as UNS 44627 stainless steel. In certain embodiments, the article of manufacture is a fuel cell interconnect for a solid oxide fuel cell.

  16. High power ferrite microwave switch

    NASA Technical Reports Server (NTRS)

    Bardash, I.; Roschak, N. K.

    1975-01-01

    A high power ferrite microwave switch was developed along with associated electronic driver circuits for operation in a spaceborne high power microwave transmitter in geostationary orbit. Three units were built and tested in a space environment to demonstrate conformance to the required performance characteristics. Each unit consisted of an input magic-tee hybrid, two non-reciprocal latching ferrite phase shifters, an out short-slot 3 db quadrature coupler, a dual driver electronic circuit, and input logic interface circuitry. The basic mode of operation of the high power ferrite microwave switch is identical to that of a four-port, differential phase shift, switchable circulator. By appropriately designing the phase shifters and electronic driver circuits to operate in the flux-transfer magnetization mode, power and temperature insensitive operation was achieved. A list of the realized characteristics of the developed units is given.

  17. Small high directivity ferrite antennas

    NASA Astrophysics Data System (ADS)

    Wright, T. M. B.

    A centimeter-wavelength antenna of millimetric dimensions, which uses the intrinsic angular sensitivity of ferrites, is described, with an emphasis on the modification of the material's permeability. The construction of both the ferrite film lens antenna and the ferrite film cassegrain antenna are detailed; both can be devised in a number of configurations for appropriate beam positioning and rf filtering. The antenna design, discussed primarily in the context of smart missiles, electronic warfare, and satellite systems, presents the possibility of magnetically switching between the transmit and receive modes within the antenna structure itself. Finally, it is noted that for a simple 2-dipole array the angular resolution can be two orders of magnitude higher than with the conventional techniques.

  18. Ferrite morphology and variations in ferrite content in austenitic stainless steel welds

    SciTech Connect

    David, S.A.; Hanzelka, S.E.; Haltom, C.P.

    1981-07-01

    Four distinct ferrite morphologies have been identified in type 308 stainless steel multipass welds: vermicular, lacy, acicular, and globular. The first three ferrite types are related to transformations following solidification and the fourth is related to the shape instability of the residual ferrite. An earlier study showed that most of the ferrite observed in austenitic stainless steel welds contaning a duplex structure may be identified as residual primary ferrite resulting from incomplete delta ..-->.. ..gamma.. transformation during solidification and/or residual ferrite after Widmanstaetten austenite precipitation in primary ferrite. These modes of ferrite formation can be used to explain observed ferrite morphologies in austenitic stainless steel welds. Variations in ferrite content within the weld were related to weld metal composition, ferrite morphology, and dissolution of ferrite resulting from thermal cycles during subsequent weld passes. An investigation of the type 308 stainless steel filler metal solidified over cooling rates ranging from 7 to 1600/sup 0/C/s showed that the cooling rate of the weld metal within the freezing range of the alloy affects the amount of ferrite in the microstructure very litte. However, the scale of the solidification substructure associated with various solidification rates may influence the ferrite dissolution kinetics.

  19. Ferrite morphology and variations in ferrite content in austenitic stainless steel welds

    SciTech Connect

    David, S.A.

    1981-04-01

    Four distinct ferrite morphologies have been identified in Type 308 stainless steel multipass welds: vermicular, lacy, acicular, and globular. The first three ferrite types are related to transformations following solidfication and the fourth is related to the shape instability of the residual ferrite. An earlier study showed that most of the ferrite observed in austenitic stainless steel welds containing a duplex structure may be identified as residual primary ferrite resulting from incomplete delta ..-->.. ..gamma.. transformation during solidification and/or residual ferrite after Widmanstatten austenite precipitation in primary ferrite. These modes of ferrite formation can be used to explain observed ferrite morphologies in austenitic stainless steel welds. Variations in ferrite content within the weld were also related to weld metal composition, ferrite morphology, and dissolution of ferrite resulting from thermal cycles during subsequent weld passes. An investigation of the Type 308 stainless steel filler metal solidified over cooling rates ranging from 7 to 1600/sup 0/C/s (44.6 to 2912/sup 0/F/s) showed that the cooling rate of the weld metal within the freezing range of the alloy affects the amount of ferrite in the microstructure very little. However, the scale of the solidification substructure associated with various solidification rates may influence the ferrite dissolution kinetics.

  20. Combined use of FLUKA and MCNP-4A for the Monte Carlo simulation of the dosimetry of 10B neutron capture enhancement of fast neutron irradiations.

    PubMed

    Pignol, J P; Cuendet, P; Brassart, N; Fares, G; Colomb, F; M'Bake Diop, C; Sabattier, R; Hachem, A; Prevot, G

    1998-06-01

    Boron neutron capture enhancement (BNCE) of the fast neutron irradiations use thermal neutrons produced in depth of the tissues to generate neutron capture reactions on 10B within tumor cells. The dose enhancement is correlated to the 10B concentration and to thermal neutron flux measured in the depth of the tissues, and in this paper we demonstrate the feasibility of Monte Carlo simulation to study the dosimetry of BNCE. The charged particle FLUKA code has been used to calculate the primary neutron yield from the beryllium target, while MCNP-4A has been used for the transport of these neutrons in the geometry of the Biomedical Cyclotron of Nice. The fast neutron spectrum and dose deposition, the thermal flux and thermal neutron spectrum in depth of a Plexiglas phantom has been calculated. The thermal neutron flux has been compared with experimental results determined with calibrated thermoluminescent dosimeters (TLD-600 and TLD-700, respectively, doped with 6Li or 7Li). The theoretical results were in good agreement with the experimental results: the thermal neutron flux was calculated at 10.3 X 10(6) n/cm2 s1 and measured at 9.42 X 10(6) n/cm2 s1 at 4 cm depth of the phantom and with a 10 cm X 10 cm irradiation field. For fast neutron dose deposition the calculated and experimental curves have the same slope but different shape: only the experimental curve shows a maximum at 2.27 cm depth corresponding to the build-up. The difference is due to the Monte Carlo simulation which does not follow the secondary particles. Finally, a dose enhancement of, respectively, 4.6% and 10.4% are found for 10 cm X 10 cm or 20 cm X 20 cm fields, provided that 100 micrograms/g of 10B is loaded in the tissues. It is anticipated that this calculation method may be used to improve BNCE of fast neutron irradiations through collimation modifications. PMID:9650176