High temperature tensile testing of modified 9Cr-1Mo after irradiation with high energy protons

International Nuclear Information System (INIS)

Toloczko, M.B.; Hamilton, M.L.; Maloy, S.A.

2003-01-01

This study examines the effect of tensile test temperatures ranging from 50 to 600 deg. C on the tensile properties of a modified 9Cr-1Mo ferritic steel after high energy proton irradiation at about 35-67 deg. C to doses from 1 to 3 dpa and 9 dpa. For the specimens irradiated to doses between 1 and 3 dpa, it was observed that the yield strength and ultimate strength decreased monotonically as a function of tensile test temperature, whereas the uniform elongation (UE) remained at approximately 1% for tensile test temperatures up to 250 deg. C and then increased for tensile test temperatures up to and including 500 deg. C. At 600 deg. C, the UE was observed to be less than the values at 400 and 500 deg. C. UE of the irradiated material tensile tested at 400-600 deg. C was observed to be greater than the values for the unirradiated material at the same temperatures. Tensile tests on the 9 dpa specimens followed similar trends

An automated tensile machine for small specimens heavily neutron irradiated in FFTF/MOTA

International Nuclear Information System (INIS)

Kohyama, Akira; Sato, Shinji; Hamada, Kenichi

1993-01-01

The objective of this work is to develop a fully automated tensile machine for post-irradiation examination (PIE) of Fast Flux Test Facility (FFTF)/Materials Open Test Assembly (MOTA) irradiated miniature tension specimens. The anticipated merit of the automated tensile machine is to reduce damage to specimens during specimen handling for PIE and to reduce exposure to radioactive specimens. This machine is designed for testing at elevated temperatures, up to 873 K, in a vacuum or in an inert gas environment. Twelve specimen assemblies are placed in the vacuum chamber that can be tested successively in a fully automated manner. A unique automated tensile machine for the PIE of FFTF/MOTA irradiated specimens, the Monbusho Automated Tensile Machine (MATRON) consists of a test frame with controlling units and an automated specimen-loading apparatus. The qualification of the test frame has been completed, and the results have satisfied the machine specifications. The capabilities of producing creep and relaxation data have been demonstrated for Cu, Al, 316SS, and ferritic steels. The specimen holders for the three-point bending test and the small bulge test (small punch test; SP test) were also designed and produced

Tensile properties of neutron irradiated solid HIP 316L(N). ITER Task T214, NET deliverable GB6 ECN-5

International Nuclear Information System (INIS)

Van Osch, E.V.; Tjoa, G.L.; Boskeljon, J.; Van Hoepen, J.

1998-05-01

The tensile properties of neutron irradiated Hot Isostatically Pressed (HIP) joints of type 316L(N) stainless steel (heat PM-130) have been measured. Cylindrical tensile test specimens of 4 mm diameter were irradiated in the High Flux Reactor (HFR) in Petten, The Netherlands, simulating the first wall conditions by a combination of high displacement damage with proportional amounts of helium. The solid HIP specimens were irradiated up to a target dose level of 5 dpa at a temperature of 550K. The damage levels realized range from 3.0 to 4.1 dpa, with helium contents up to 38 appm. Post irradiation testing temperatures ranged from 300 to 700K. The report contains the experimental conditions and summarises the results, which are given in terms of engineering stresses and strains and reduction of area. The main conclusions are that the unirradiated solid-HIP material is very soft, assumingly due to the relatively large grain size. Neutron irradiation induces both hardening and reduction of ductility, similar to the behaviour of 316L(N) plate. No failures related to debonding were observed for the tests of the unirradiated samples, however one of eight tested irradiated specimens fractured in the HIP joint, showing a flat fracture surface and a low reduction of area. 6 refs

Empirical relations for tensile properties of austenitic stainless steels irradiated in mixed-spectrum reactors

International Nuclear Information System (INIS)

Grossbeck, M.L.

1991-01-01

An assessment has been made of available tensile property data relevant to the design of fusion reactors, especially near term devices expected to operate at lower temperatures than power reactors. Empirical relations have been developed for the tensile properties as a functions of irradiation temperature for neutron exposures of 10-15, 20, 30, and 50 dpa. It was found that yield strength depends little on the particular austenitic alloy and little on the helium concentration. Strength depends upon initial condition of the alloy only for exposures of less than 30 dpa. Uniform elongation was found to be more sensitive to alloy and condition. It was also more sensitive than strength to helium level. However, below 500deg C, helium only appeared to have an efect at 10-15 dpa. At higher temperatures, helium embrittlement was apparent, and its threshold temperature decreased with increasing neutron exposure level. (orig.)

High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

Energy Technology Data Exchange (ETDEWEB)

Roy, R B; Solly, B

1966-12-15

The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

Effects of gamma-ray irradiation on tensile properties of ultradrawn polyethylene

International Nuclear Information System (INIS)

Iida, Shozo; Sakami, Hiroshi

1977-01-01

The deformation of ultradrawn polyethylene was previously shown that crystalline chains were pulled out by the tension applied to tie chains which connected crystal blocks. This paper deals with the effects of γ-ray irradiation on crosslinking which prevent crystalline chains being pulled out and to improve the tensile properties. The tensile strength of high density polyethylene, drawn by a factor of 40, increased from 73 to 113 kg/mm 2 at 20 0 C and from 13 to 42 kg/mm 2 at 80 0 C with increasing irradiation dose from zero to 100 Mrad. The tensile elongation, the residual strain measured by cyclic strain test, and the rate of stress decrease by the stress relaxation measurement diminished with increasing irradiation dose. These facts showed the existence of preventive effects by crosslinks on pulling. The stress-strain relation of crosslinked polymer was calculated thermodynamically from the melting of crystalline chains accompanied by the sliding of chains, assuming that the sliding of crystalline chains was brought about by an unbalance of the tension applied to tie chains with which both sides of crystalline chains were connected. The equation of stress was derived; stress increased with increasing strain and was proportional to the Gibbs' free energy of fusion. The observed stress-strain relations obeyed the above mentioned equation. (auth.)

Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Shiba, K.; Ioka, I.; Jitsukawa, S.; Hamada, A.; Hishinuma, A. [and others

1996-10-01

Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400{degrees}C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small for not only base metal specimens but also for the weld joint and the weld metal specimens.

Tensile mechanical properties of a stainless steel irradiated up to 19 dpa in the Swiss spallation neutron source

Energy Technology Data Exchange (ETDEWEB)

Saito, Shigeru, E-mail: saito.shigeru@jaea.go.jp [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kikuchi, Kenji [Ibaraki Univ., iFRC, Tokai-mura, Ibaraki-ken 319-1106 (Japan); Hamaguchi, Dai [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Usami, Kouji; Endo, Shinya; Ono, Katsuto; Matsui, Hiroki [JAEA, Dept. of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, Masayoshi [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Yong [PSI, Spallation Source Division, Villigen PSI (Switzerland)

2012-12-15

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS), post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens was carried out. The specimens tested in this study were made from the austenitic steel Japan primary candidate alloy (JPCA). The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580 MeV, irradiation temperatures ranged from 100 to 430 Degree-Sign C, and displacement damage levels ranged from 7.1 to 19.5 dpa. Tensile tests were performed in air at room temperature (RT), 250 Degree-Sign C and 350 Degree-Sign C. Fracture surface observation after the tests was done by Scanning electron microscope (SEM). Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission neutron irradiated data. At the higher temperatures, 250 Degree-Sign C and 350 Degree-Sign C, the extra hardening was not observed. Degradation of ductility bottomed around 10 dpa, and specimens kept their ductility until 19.5 dpa. All specimens fractured in ductile manner.

Evolution of cleared channels in neutron-irradiated pure copper as a function of tensile strain

DEFF Research Database (Denmark)

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

2004-01-01

Tensile specimens of pure copper were neutron irradiated at similar to323 K to a displacement dose of 0.3 dpa (displacement per atom). Five irradiated specimens were tensile tested at 300 K, but four of the specimens were stopped at specific strains -just before the yield point at similar to90......% of the macroscopic yield, at 1.5% and 5% elongation, and near the ultimate tensile strength at 14.5% elongation, with the 5th specimen tested to failure (e(T) = 22%). SEM and TEM characterization of the deformed specimens revealed that the plastic strain was confined primarily to the 'cleared' channels only...

Increased Tensile Strength of Carbon Nanotube Yarns and Sheets through Chemical Modification and Electron Beam Irradiation

Science.gov (United States)

Miller, Sandi G.; Williams, Tiffany S.; Baker, James S.; Sola, Francisco; Lebron-Colon, Marisabel; McCorkle, Linda S.; Wilmoth, Nathan G.; Gaier, James; Chen, Michelle; Meador, Michael A.

2014-01-01

The inherent strength of individual carbon nanotubes offers considerable opportunity for the development of advanced, lightweight composite structures. Recent work in the fabrication and application of carbon nanotube (CNT) forms such as yarns and sheets has addressed early nanocomposite limitations with respect to nanotube dispersion and loading; and has pushed the technology toward structural composite applications. However, the high tensile strength of an individual CNT has not directly translated to macro-scale CNT forms where bulk material strength is limited by inter-tube electrostatic attraction and slippage. The focus of this work was to assess post processing of CNT sheet and yarn to improve the macro-scale strength of these material forms. Both small molecule functionalization and e-beam irradiation was evaluated as a means to enhance tensile strength and Youngs modulus of the bulk CNT material. Mechanical testing results revealed a tensile strength increase in CNT sheets by 57 when functionalized, while an additional 48 increase in tensile strength was observed when functionalized sheets were irradiated; compared to unfunctionalized sheets. Similarly, small molecule functionalization increased yarn tensile strength up to 25, whereas irradiation of the functionalized yarns pushed the tensile strength to 88 beyond that of the baseline yarn.

Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, E.; Schneider, H.-C.; Möslang, A.

2014-01-01

Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects

Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

Energy Technology Data Exchange (ETDEWEB)

Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu; Schneider, H.-C., E-mail: hans-christian.schneider@kit.edu; Möslang, A., E-mail: anton.moeslang@kit.edu

2014-12-15

Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects.

Tensile properties of several 800 MeV proton-irradiated bcc metals and alloys

International Nuclear Information System (INIS)

Brown, R.D.; Wechsler, M.S.; Tschalar, C.

1987-01-01

A spallation neutron source for the 600-MeV proton accelerator facility at the Swiss Institute for Nuclear Research (SIN) consists of a vertical cylinder filled with molten Pb-Bi. The proton beam enters the cylinder, passing upward through a window in contact with the Pb-Bi eutectic liquid that must retain reasonable strength and ductility upon irradiation at about 673 K to fluence of about 1 x 10/sup 25/ protons/m/sup 2/. Investigations are underway at the 800-MeV proton accelerator at the Los Alamos Meson Physics Facility (LAMPF) to test the performance of candidate SIN window materials under appropriate conditions of temperature, irradiation, and environment. Based on considerations of chemical compatibility with molten Pb-Bi, as well as interest in identifying fundamental radiation damage mechanisms, Fe, Ta, Fe-2.25Cr-1Mo, and Fe-12Cr-1Mo(HT-9) were chosen as candidate materials. Sheet tensile samples, 0.5-mm thick, of the four materials were fabricated and heat treated. The samples were sealed inside capsules containing Pb-Bi and were proton-irradiated at LAMPF to two fluences, 4.8 and 54 x 10/sup 23/ p/m/sup 2/. The beam current was approximately equal to the 1 mA anticipated for the upgraded SIN accelerator. The power deposited by the proton beam in the capsules was sufficient to maintain sample temperatures of about 673 K. Post-irradiation tensile tests were conducted at room temperature at a strain rate of 9 x 10/sup -4/s/sup -1/. The yield and ultimate strengths increased upon irradiation in all materials, while the ductility decreased, as indicated by the uniform strain. The pure metals, Ta and Fe, exhibited the greatest radiation hardening and embrittlement. The HT-9 alloy showed the smallest changes in strength and ductility. The increase in strength following irradiation is discussed in terms of a dispersed-barrier hardening model, for which the barrier sizes and formation cross sections are calculated

Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

2015-01-01

Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

International Nuclear Information System (INIS)

Grossbeck, M.L.; Wiffen, F.W.

1985-01-01

The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs

Effect of ion irradiation on tensile ductility, strength and fictive temperature in metallic glass nanowires

International Nuclear Information System (INIS)

Magagnosc, D.J.; Kumar, G.; Schroers, J.; Felfer, P.; Cairney, J.M.; Gianola, D.S.

2014-01-01

Ion irradiation of thermoplastically molded Pt 57.5 Cu 14.3 Ni 5.7 P 22.5 metallic glass nanowires is used to study the relationship between glass structure and tensile behavior across a wide range of structural states. Starting with the as-molded state of the glass, ion fluence and irradiated volume fraction are systematically varied to rejuvenate the glass, and the resulting plastic behavior of the metallic glass nanowires probed by in situ mechanical testing in a scanning electron microscope. Whereas the as-molded nanowires exhibit high strength, brittle-like fracture and negligible inelastic deformation, ion-irradiated nanowires show tensile ductility and quasi-homogeneous plastic deformation. Signatures of changes to the glass structure owing to ion irradiation as obtained from electron diffraction are subtle, despite relatively large yield strength reductions of hundreds of megapascals relative to the as-molded condition. To reconcile changes in mechanical behavior with glass properties, we adapt previous models equating the released strain energy during shear banding to a transit through the glass transition temperature by incorporating the excess enthalpy associated with distinct structural states. Our model suggests that ion irradiation increases the fictive temperature of our glass by tens of degrees – the equivalent of many orders of magnitude change in cooling rate. We further show our analytical description of yield strength to quantitatively describe literature results showing a correlation between severe plastic deformation and hardness in a single glass system. Our results highlight not only the capacity for room temperature ductile plastic flow in nanoscaled metallic glasses, but also processing strategies capable of glass rejuvenation outside of the realm of traditional thermal treatments

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

International Nuclear Information System (INIS)

Billone, M.C.

1998-01-01

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

Postirradiation tensile properties of Mo and Mo alloys irradiated with 600 MeV protons

International Nuclear Information System (INIS)

Mueller, G.V.; Gavillet, D.; Victoria, M.; Martin, J.L.

1994-01-01

Tensile specimens of pure Mo and Mo-5 Re, Mo-41 Re and TZM alloys have been irradiated with 600 MeV protons in the PIREX facility at 300 and 660 K to 0.5 dpa. Results of the postirradiation tensile testing show a strong radiation hardening and a severe loss of ductility for all the materials tested at room temperature. ((orig.))

Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

Energy Technology Data Exchange (ETDEWEB)

Novak, J [Czech Nuclear Society, Prague (Czech Republic)

1994-12-31

Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs.

Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

International Nuclear Information System (INIS)

Novak, J.

1993-01-01

Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs

Experimental assessments of notch ductility and tensile strength of stainless steel weldments after 1200C neutron irradiation

International Nuclear Information System (INIS)

Hawthorne, J.R.; Menke, B.H.; Awadalla, N.G.; O'Kula, K.R.

1986-01-01

The Charpy-V (C/sub v/) properties of AISI 300 series stainless steel plate, weld, and weld heat-affected zone (HAZ) materials from commercial production weldments in 406-mm-diameter pipe (12.7-mm wall) were investigated in unirradiated and irradiated conditions. Weld and HAZ tensile properties were also assessed in the two conditions. The plates and weld filler wires represent different steel melts; the welds were produced using the multipass MIG process. Weldment properties in two test orientations were evaluated. Specimens were irradiated in the UBR reactor to 1 x 10 20 n/cm 2 , E >0.1 MeV in a controlled temperature assembly. Specimen tests were performed at 25 0 C and 125 0 C. The radiation-induced reductions in C/sub v/ energy absorption at 25 0 C were about 42 percent for the weld and HAZ materials evaluated. A trend of energy increase with temperature was observed. The concomitant elevation in yield strength was about 53%. In contrast, the increase in tensile strength was only 16%. The postirradiation yield strength of the axial test orientation in the pipe was less than that of the circumferential test orientation. Results for the HAZ indicate that this component may be the weakest link in the weldment from a fracture resistant viewpoint

Tensile and electrical properties of unirradiated and irradiated Hycon 3HP{trademark} CuNiBe

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J.; Eatherly, W.S. [Oak Ridge National Lab., TN (United States)

1996-10-01

The unirradiated tensile properties of two different heats of Hycon 3HP{trademark} CuNiBe (HT Temper) have been measured over the temperature range of 20-500{degrees}C for longitudinal and long transverse orientations. The room temperature electrical conductivity has also been measured for both heats. Both heats exhibited a very good combination of strength and conductivity at room temperature. The strength remained relatively high at all test temperatures, with a yield strength of 420-520 MPa at 500{degrees}C. However, low levels of ductility (<5% uniform elongation) were observed at test temperatures above 200-250{degrees}C, due to flow localization adjacent to grain boundaries. Fission neutron irradiation to a dose of {approximately}0.7 dpa at temperatures between 100 and 240{degrees}C produced a slight increase in strength and a significant decrease in ductility. The measured tensile elongation increased with increasing irradiation temperature, with a uniform elongation of {approximately}3.3% observed at 240{degrees}C. The electrical conductivity decreased slightly following irradiation, due to the presence of defect clusters and Ni, Zn, Co transmutation products. The data indicate that CuNiBe alloys have irradiated tensile and electrical properties comparable or superior to CuCrZr and oxide dispersion strengthened copper at temperatures <250{degrees}C, and may be suitable for certain fusion energy structural applications.

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Experimental assessments of notch ductility and tensile strength of stainless steel weldments after 1200C neutron irradiation

International Nuclear Information System (INIS)

Hawthorne, J.R.; Menke, B.H.; Awadalla, N.G.; O'Kula, K.R.

1987-01-01

The Charpy-V (C/sub V/) properties of American Iron and Steel Institute (AISI) 300 series stainless steel plate, weld, and weld heat-affected zone (HAZ) materials from commercial production weldments in 406-mm-diameter pipe (12.7-mm wall) were investigated in unirradiated and irradiated conditions. Weld and HAZ tensile properties were also assessed in the two conditions. The plates and weld filler wires represent different steel melts; the welds were produced using the multipass metal inert gas (MIG) process. Weldment properties in two test orientations were evaluated. Specimens were irradiated in a light water cooled and moderated reactor to 1 x 10/sup 20/ n/cm/sup 2/, E > 0.1 MeV, using a controlled temperature assembly. Specimen tests were performed at 25 and 125 0 C. The radiation-induced reductions in C/sub V/ energy absorption at 25 0 C were about 42% for the weld and the HAZ materials evaluated. A trend of energy increase with temperature was observed. The concomitant elevation in yield strength was about 53%. The increase in tensile strength in contrast was only 16%. The postirradiation yield strength of the axial test orientation in the pipe was less than that of the circumferential test orientation. Results for the HAZ indicate that this component may be the weakest link in the weldment from a fracture resistance viewpoint

Tensile and low cycle fatigue properties of EUROFER97-steel after 16.3 dpa neutron irradiation at 523, 623 and 723 K

Energy Technology Data Exchange (ETDEWEB)

Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu; Möslang, A., E-mail: anton.moeslang@kit.edu; Schneider, H.-C., E-mail: hans-christian.schneider@kit.edu

2013-11-15

Neutron-irradiated specimens of the reduced-activation tempered martensitic steel EUROFER97 were tested by tensile and low cycle conditions to detect the impact of irradiation on strength and lifetime. The irradiation temperature ranged from 523 to 723 K with an accumulated dose of up to 16.3 dpa. Tensile tests revealed a significant irradiation-induced hardening below 673 K with a peak of ∼430 MPa at 573 K but none was seen at 723 K, as expected. Despite the significant irradiation-induced reduction of uniform elongation, the total elongation is only reduced by about 50% below 673 K. Post-irradiation strain-controlled fatigue tests have been carried out at T{sub irrad} = T{sub test} = 523, 623 and 723 K. Pronounced cyclic softening was observed in all specimens. At 623 and 723 K, neutron irradiation had no effect on fatigue life within the data scatter. A significant lifetime increase has been observed at T{sub irrad} = T{sub test} = 523 K that advances with decreasing stress amplitude Δε (1% → 0.5%) up to a factor of ten. Scanning electron microscopy (SEM) analysis revealed ductile fracture and fatigue striations on the fracture surfaces. After push–pull fatigue testing, transmission electron microscopy (TEM) investigations showed the typical sub-cell formation, even at T{sub irrad} = T{sub test} = 523 K.

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

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

1997-08-01

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

Tensile and shear fracture behavior of fiber reinforced plastics at 77K irradiated by various radiation sources

International Nuclear Information System (INIS)

Humer, K.; Weber, H.W.; Tschegg, E.K.; Gerstenberg, H.

1993-08-01

Influence of radiation damage (gamma, electron, neutron) on mechanical properties of fiber reinforced plastics (FRPs) has been investigated. Different types of FRPs (two or three dimensional E-, S- or T-glass fiber reinforcement, epoxy or bismaleimide resin) have been irradiated at room temperature with 2 MeV electrons and 6O Co γ-rays up to 1.8 x 1 0 8 Gy as well as with different reactor spectra up to a fast neutron fluence of 5 x lO 22 m -2 (E > 0.1 MeV). Tensile and intralaminar shear tests were carried out on the irradiated samples at 77 K. Some samples were irradiated at 5 K and tested at 77 K with and without an annealing cycle to room temperature. Results on the influence of these radiation conditions and of warm-up cycles on the mechanical properties of FRPs are compared and discussed

Hardness distribution and tensile properties in an electron-beam-welded F82H irradiated in HFIR

International Nuclear Information System (INIS)

Hashimoto, N.; Oka, H.; Muroga, T.; Kimura, A.; Sokolov, M.A.; Yamamoto, T.

2014-01-01

F82H-IEA and its EB-weld joint were irradiated at 573 and 773 K up to 9.6 dpa in the HFIR and the irradiation effect on its mechanical properties and microstructure were investigated. A hardness profile across the weld joint before irradiation showed the hardness in transformed region (TR) was high and especially that in the edge of TR was the highest (high hardness region: HHR) compared to base metal. This hardness distribution corresponds to grain size distribution. After irradiation, hardening in HHR was small compared to other region in the sample. In tensile test, the amount of hardening in yield strength and ultimate tensile strength of F82H EB-weld joint was almost similar to that of F82H-1EA but the fracture position of EB-weld joint was at the boundary of TR and BM. Therefore, the TR/BM boundary is the structural weak point in F82H EB-weld joint after irradiation. As the plastic instability was observed, the dislocation channeling deformation can be expected though the dislocation channel was not observed in this study. (author)

Tensile and shear fracture behavior of fiber reinforced plastics at 77K irradiated by various radiation sources

Energy Technology Data Exchange (ETDEWEB)

Humer, K.; Weber, H.W. [Atominstitut der Oesterreichischen Hochschulen, Vienna (Austria); Tschegg, E.K. [Technische Univ., Vienna (Austria). Inst. fuer Angewandte und Technische Physik; Egusa, Shigenori [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment; Birtcher, R.C. [Argonne National Lab., IL (United States); Gerstenberg, H. [Technische Univ. Muenchen, Garching (Germany). Fakultaet fuer Physik

1993-08-01

Influence of radiation damage (gamma, electron, neutron) on mechanical properties of fiber reinforced plastics (FRPs) has been investigated. Different types of FRPs (two or three dimensional E-, S- or T-glass fiber reinforcement, epoxy or bismaleimide resin) have been irradiated at room temperature with 2 MeV electrons and {sup 6O}Co {gamma}-rays up to 1.8 {times} 1 0{sup 8} Gy as well as with different reactor spectra up to a fast neutron fluence of 5 {times} lO{sup 22} m{sup {minus}2} (E > 0.1 MeV). Tensile and intralaminar shear tests were carried out on the irradiated samples at 77 K. Some samples were irradiated at 5 K and tested at 77 K with and without an annealing cycle to room temperature. Results on the influence of these radiation conditions and of warm-up cycles on the mechanical properties of FRPs are compared and discussed.

Influence of tensile stress on cavity growth in nickel under helium irradiation

International Nuclear Information System (INIS)

Kusanagi, Hideo; Hide, Koichiro; Takaku, Hiroshi

1989-01-01

The influence of tensile stress on cavity behavior in pure nickel under helium irradiation was investigated by in-situ observation using the transmission electron microscope (TEM) in which an ion gun is installed. Specimens were irradiated at 500 0 C with 20 keV helium in the TEM. The dose rate was about 10 14 He/cm 2 s, and the angle between the helium beam and the normal direction of the specimens was about 60 0 . The damage rate estimated by the E-DEP-1 code was about 0.6x10 -3 dpa/s at its peak position. The main results are as follows: (1) cavity nucleation was accelerated by applying tensile stress, and cavity size in stressed specimens was several times larger than that in stress-free specimens; (2) cavity density in the stressed specimen increased more rapidly than in the stress-free specimen, and then decreased by cavity coalescences; (3) depth of cavity nucleation in the stress-free specimen was about 160 nm, while that in the stressed specimen was about 320 nm; that is, cavities nucleated in deeper regions in the stressed specimen than in the stress-free specimen. This result indicates that helium atoms and vacancies can migrate into the deeper region by applying tensile stress. (4) The experimental results obtained in this study can be explained qualitatively by the mechanism that mobile dislocations drag He-V complexes to the deeper region. This implies that there are similar phenomena in the case of compressive stress. (orig.)

Proton irradiation effects on tensile and bend-fatigue properties of welded F82H specimens

Energy Technology Data Exchange (ETDEWEB)

Saito, S., E-mail: saito.shigeru@jaea.go.j [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Kikuchi, K.; Hamaguchi, D. [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Usami, K.; Ishikawa, A.; Nishino, Y.; Endo, S. [JAEA Tokai, Department of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, M. [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Y. [PSI, Spallation Source Division, 5232 Villigen PSI (Switzerland)

2010-03-15

In several institutes, research and development for an accelerator-driven transmutation system (ADS) have been progressed. Ferritic/martensitic (FM) steels are the candidate materials for the beam window of ADS. To evaluate of the mechanical properties of the irradiated materials, the post irradiation examination (PIE) work of the SINQ (Swiss spallation neutron source) target irradiation program (STIP) specimens was carried out at JAEA. In present study, the results of PIE on FM steel F82H and its welded joint have been reported. The present irradiation conditions of the specimens were as follows: proton energy was 580 MeV. Irradiation temperatures were ranged from 130 to 380 deg. C, and displacement damage level was ranged from 5.7 to 11.8 dpa. The results of tensile tests performed at 22 deg. C indicated that the irradiation hardening occurred with increasing the displacement damage up to 10.1 dpa at 320 deg. C irradiation. At higher dose (11.8 dpa) and higher temperature (380 deg. C), irradiation hardening was observed, but degradation of ductility was relaxed in F82H welded joint. In present study, all specimens kept its ductility after irradiation and fractured in ductile manner. The results on bend-fatigue tests showed that the fatigue life (N{sub f}) of F82H base metal irradiated up to 6.3 dpa was almost the same with that of unirradiated specimens. The N{sub f} of the specimens irradiated up to 9.1 dpa was smaller than that of unirradiated specimens. Though the number of specimen was limited, the N{sub f} of F82H EB (15 mm) and EB (3.3 mm) welded joints seemed to increase after irradiation and the fracture surfaces of the specimens showed transgranular morphology. While F82H TIG welded specimens were not fractured by 10{sup 7} cycles.

Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

International Nuclear Information System (INIS)

Le Saux, M.

2008-01-01

The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

Irradiation testing of stainless steel plate material and weldments. Report on ITER Task T14, Part B. Tensile properties after 0.5 and 5 dpa at 350 and 500 K

International Nuclear Information System (INIS)

Rensman, J.W.; Boskeljon, J.; Horsten, M.G.; De Vries, M.I.

1997-10-01

The tensile properties of unirradiated and neutron irradiated type 316L(N)-SPH stainless steel plate, EB weldments, 16-8 TIG-weldments, and full 16-8 TIG-deposits have been measured. Miniature 4 mm diameter test specimens of the European Reference Heat 1 and 2 (ERH), and 4 mm and some 8 mm diameter specimens of the weldments mentioned above, were irradiated in the High Flux Reactor (HFR) in Petten, The Netherlands, simulating the first wall conditions by a combination of high displacement damage with high amounts of helium. The irradiation conditions were 0.5 and 5 displacements per atom (dpa) at 350K and 0.5 and 5 dpa at 500K. Testing temperatures ranged from 300K to 850K. This work was performed as part of the European Fusion Technology Programme for ITER as 'Irradiation testing of stainless steel' The report contains the experimental conditions and summarises the results. The tensile properties of the unirradiated ERH's 1 and 2 plate materials were found to differ slightly but significantly: ERH2 has a lower UTS, but higher yield strength and ductility than ERH1. The plate materials have lower yield strength in the unirradiated condition than all of the weldments (EB, TIG-weld and TIG-deposit), accompanied by a higher ductility of the plate materials. When irradiated at 350K the differences in strength between the plate and weld materials decrease, but the ductility of the plate remains higher than that of the weldments. A saturation of irradiation damage has taken place already at about 0.5 dpa. When irradiated at 500K the plate material continuously hardens up to 5 dpa, where it has lost all uniform plastic ductility. The weldments show similar but less dramatic hardening and loss of ductility as the plate material for both irradiation conditions. 54 figs., 17 tabs., 21 refs

Microstructure and mechanical properties of neutron irradiated OFHC-copper before and after post-irradiation annealing

International Nuclear Information System (INIS)

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

2001-02-01

Tensile specimens of OFHC-copper were irradiated with fission neutrons in the DR-3 reactor at Risoe National Laboratory at 100 deg. C to different displacement dose levels in the range of 0.01 to 0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas other were given a post-irradiation annealing at 300 deg. C for 50 h and subsequently tested at 100 deg. C. Transmission electron microscopy was used to characterize the microstructure of specimens in the as-irradiation as well as irradiation and annealed conditions both before and after tensile deformation. The results show that while the interstitial loop microstructure coarsens with irradiation dose, no significant changes are observed in the population of stacking fault tetrahedra. The results also illustrates that the post-irradiation annealing leads to only a partial recovery and that the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the problem of yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade induced source hardening (CISH) and the dispersed barrier hardening (DBH) models. Both technological and scientific implications of these results are considered. (au)

Status of automated tensile machine

International Nuclear Information System (INIS)

Satou, M.; Hamilton, M.L.; Sato, S.; Kohyama, A.

1992-01-01

The objective of this work is to develop the Monbusho Automated Tensile machine (MATRON) and install and operate it at the Pacific Northwest Laboratory (PNL). The machine is designed to provide rapid, automated testing of irradiated miniature tensile specimen in a vacuum at elevated temperatures. The MATRON was successfully developed and shipped to PNL for installation in a hot facility. The original installation plan was modified to simplify the current and subsequent installations, and the installation was completed. Detailed procedures governing the operation of the system were written. Testing on irradiated miniature tensile specimen should begin in the near future

Degradation in tensile properties of aromatic polymers by electron beam irradiation

International Nuclear Information System (INIS)

Sasuga, T.; Hayakawa, N.; Yoshida, K.; Hagiwara, M.

1985-01-01

Electron beam irradiation effects of ten kinds of polymers containing various aromatic rings linked by functional groups in the main chain (aromatic polymer) were studied with reference to change in tensile properties. The polymers studied were polyimides 'Kapton H', and 'UPILEX', polyetherimide 'ULTEM', polyamides 'A-Film' and 'APH-50 (nomex type paper)', poly-ether-ether-ketone 'PEEK', polyarylate 'U-Polymer', polysulphones 'Udel-Polysulphone' and 'PES', and modified poly(phenylene oxide) 'NORYL'. Irradiation was carried out by use of electron beam at a dose rate of 5 x 10 3 Gy s -1 at room temperature. The elongation at break was the most severely influenced by the irradiation and it decreased with increasing dose. The order of radiation resistivity which was evaluated from the dose required for the elongation to become 50% and 20% of the initial value was as follows: Polyimide > PEEK > polyamide > polyetherimide > polyarylate > polysulphone, poly(phenylene oxide). Based on the above experimental results, an order is proposed for the radiation stability of the aromatic repeating units composing the main chain. (author)

Tensile properties and bend ductility of (Fe,Ni)3V long-range-ordered alloys after irradiation in HFIR

International Nuclear Information System (INIS)

Braski, D.N.

1984-01-01

The objective of this work was to determine the effect of neutron irradiation on the tensile properties and bend ductility of (Fe,Ni) 3 V long-range-ordered (LRO) alloys. Several (Fe,Ni) 3 V LRO alloys were irradiated in HFIR-CTR-42 and -43 at 400 to 600 0 C, to approximately 10 dpa and approximately 1000 at. ppm He. Additions of cerium or carbon and the use of cold-worked microstructures did not improve the embrittlement resistance of the LRO alloys. The LRO-37-5RS alloy, with a microstructure produced by rapid solidification, exhibited the highest ductilities, and further study of the RS microstructure is warranted. The correlation between bend ductility and tensile ductility was poor

Ion irradiation effects on tensile properties of carbon fibres

International Nuclear Information System (INIS)

Kurumada, A.; Ishihara, M.; Baba, S.; Aihara, J.

2004-01-01

Carbon/carbon composite materials have high thermal conductivity and excellent mechanical properties at high temperatures. They have been used as structural materials at high temperatures in fission and experimental fusion reactors. The changes in the microstructures and the mechanical properties due to irradiation damage must be measured for the safety design and the life assessment of the materials. The purpose of this study is to obtain a basic knowledge of the development of new carbon composite materials having high thermal conductivity and excellent resistance to irradiation damage. Five kinds of carbon fibres were selected, including a vapour growth carbon fibre (VGCF; K1100X), a polyacrylonitrile-based fibre (PAN; M55JB by Toray Corp.), two meso-phase pitch-based fibres (YS-15-60S and YS-70-60S by Nippon Graphite Fiber Corp.) and a pitch-based fibre (K13C2U by Mitsubishi Chemical Co.). They were irradiated by high-energy carbon, nickel and argon ions. Irradiation damages in the carbon fibres are expected to be uniform across the cross-section, as the diameters of the carbon fibres are about 20 μm and are sufficiently smaller than the ranges of ions. The cross-sectional areas increased due to ion irradiation, with the exception of the K1100X of VGCF. One of the reasons for the increases is the swelling of carbon basal planes due to lattice defects in the graphite interlayer. The tensile strengths and the Young's moduli decreased due to ion irradiation except for the K1100X of VGCF and the YS-15-60S of meso-phase pitch-based fibres. One of the reasons for the decreases is thought to be that the microstructures of carbon fibres are damaged in the axial direction, as ions were irradiated vertically with respect to the longitudinal direction of carbon fibres. The results of this study indicate that the VGCF and the meso-phase pitch-based carbon fibres could be useful as reinforcement fibres of new carbon composite materials having high thermal conductivity and

Swelling and tensile properties of neutron-irradiated vanadium alloys

International Nuclear Information System (INIS)

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

1990-07-01

Vanadium-base alloys are candidates for use as structural material in magnetic fusion reactors. In comparison to other candidate structural materials (e.g., Type 316 stainless and HT-9 ferritic steels), vanadium-base alloys such as V-15Cr-5Ti and V-20Ti have intrinsically lower long-term neutron activation, neutron irradiation after-heat, biological hazard potential, and neutron-induced helium and hydrogen transmutation rates. Moreover, vanadium-base alloys can withstand a higher surface-heat, flux than steels because of their lower thermal stress factor. In addition to having these favorable neutronic and physical properties, a candidate alloy for use as structural material in a fusion reactor must have dimensional stability, i.e., swelling resistance, and resistance to embrittlement during the reactor lifetime at a level of structural strength commensurate with the reactor operating temperature and structural loads. In this paper, we present experimental results on the swelling and tensile properties of several vanadium-base alloys after irradiation at 420, 520, and 600 degree C to neutron fluences ranging from 0.3 to 1.9 x 10 27 neutrons/m 2 (17 to 114 atom displacements per atom [dpa])

Correlation of fracture toughness with tensile properties for irradiated 20% cold-worked 316 stainless steel

International Nuclear Information System (INIS)

Hamilton, M.L.; Garner, F.A.; Wolfer, W.G.

1983-08-01

A correlation has been developed which allows an estimate to be made of the toughness of austenitic alloys using more easily obtained tensile data. Tensile properties measured on 20% cold-worked AISI 316 specimens made from ducts and cladding irradiated in EBR-II were used to predict values for the plane strain fracture toughness according to a model originally developed by Krafft. Some microstructural examination is required to determine a parameter designated as the process zone size. In contrast to the frequently employed Hahn-Rosenfeld model, this model gives results which agree with recent experimental determinations of toughness performed in the transgranular failure regime

Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

International Nuclear Information System (INIS)

Klueh, R.L.; Vitek, J.M.

1984-01-01

The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

Microstructure and tensile properties of neutron-irradiated (FE061Ni039)3V ordered alloy

International Nuclear Information System (INIS)

Braski, D.N.

1982-01-01

Small tensile specimens of the (Fe 0 61 Ni 0 39 ) 3 V long-range-ordered alloy were irradiated in the ORR to 4 dpa at 523, 623, and 823 K and subsequently tested at the same respective temperatueres. The alloy remained ordered after irradiation at all three temperatures. Irradiation at 523 and 623 K increased the yield strength of the material by producing Frank loops in the microstructure and reduced the total elongation. The low strain hardening observed was attributed to planar slip and the absence of cross slip. Irradiation at 823 K embrittled the alloy. Premature failure was apparently initiated by helium bubbles on sigma phase boundaries which grew rapidly during the test to form microcracks. Fracture occurred after a microcrack propagated across grain boundaries that were weakened by helium and possible sulfur. New LRO alloys without sigma phase should perform better under neutron irradiation

Ductility and failure behaviour of both unirradiated and irradiated zircaloy-4 cladding using plane strain tensile specimens

International Nuclear Information System (INIS)

Carassou, S.; Le Saux, M.; Pizzanelli, J.P.; Rabouille, O.; Averty, X.; Poussard, C.; Cazalis, B.; Desquines, J.; Bernaudat, C.

2010-01-01

In this work, eight PST (Plan Strain Tensile) tests machined from a Zircaloy-4 (Zy-4) cladding irradiated up to 5 annual cycles have been performed at 280, 350 and 480 Celsius degrees. The specimen displacements during the tests were filmed and digitally recorded to allow the use of a Digital Image Correlation (DIC) analysis technique to experimentally determine the local strains on the outer surface of the specimens. The plane strain conditions have been verified and prevail over a wide area between the notches of the specimen, as expected from full 3D FE numerical analysis performed in support of the tests. For the first time, the location of the onset of fracture for this geometry on irradiated material has been experimentally observed: at 280 C.degrees, crack initiates in the vicinity of the notches, in an area where plane strain conditions are not fulfilled, and for a local circumferential strain value of about 5%. At 350 C. degrees and 480 C. degrees, cracks initiate at a location where plane strain conditions prevail, for circumferential strain values respectively close to 10% and greater than 50%. These results have been compared to results obtained previously by similar test on fresh and hydrided material, as well as tests performed as support to the study. At 350 C. degrees, the homogeneous 700 ppm hydrided Zy-4 and the Zy-4 irradiated during 5 annual cycles exhibit similar fracture behaviour, for both fracture hoop strain values (10%) and fracture mode (through-wall slant fracture). For the irradiated material, it has clearly been established that at 350 C. degrees, a brittle fracture occurs at the outer surface in the hydride rim. The crack propagates subsequently toward the inner surface and the notches, where final fracture occurs

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

International Nuclear Information System (INIS)

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

2001-01-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N. E-mail: bachu.singh@risoe.dk; Edwards, D.J.; Toft, P

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Science.gov (United States)

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

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risø at 100 °C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 °C whereas others were given a post-irradiation annealing treatment at 300 °C for 50 h and subsequently tested at 100 °C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Tensile properties of irradiated and fatigue exposed stainless steel DIN X 6 CrNi 1811 (similar to AISI type 304) plate and welded joints

International Nuclear Information System (INIS)

Vries, M.I. de; Schaaf, B. van der; Elen, J.D.

1979-10-01

Test specimens of plate metal and welded joints of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 n.m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High-rate (depsilon/dt = 1 s -1 ) tensile tests were performed after fatigue exposure up to various fractions of fatigue life (D) ranging from 5% to 95% at the same temperatures as the nominal temperatures of the irradiation series

Evaluation of the tensile bond strength of an adhesive system self-etching in dentin irradiated with Er:YAG laser

International Nuclear Information System (INIS)

Mello, Andrea Malluf Dabul de

2000-01-01

Since Buonocore (1955), several researchers have been seeking for the best adhesive system and treatment for the enamel and dentin surfaces. The use of the acid has been presented as one of the best techniques of dentin conditioning , because this promotes the removal of the 'smear layer and exhibition of dentinal structure, for a best penetration and micro- retention of the adhesive system. However, some conditioning methods have been appearing in the literature, for the substitution or interaction with the acid substances, as the laser. The objective of this work is to evaluate the tensile bond strength of the adhesive system self-etching' associated to a composed resin, in dentin surfaces conditioned with the Er:YAG laser. For this study, freshly extracted human teeth were used and in each one the dentinal surfaces , which were treated with three sandpapers of different granulations (120,400,600), to obtain a standard of the smear layer, before the irradiation of the laser and of the restoring procedure. After these procedures the specimens were storage in distilled water at 37 deg C for 24 hours. Soon after, they were submitted to the tensile strength test .After analyzing the results, we can concluded that the use of the Er:YAG laser can substitute the drill without the need of conditioning, when using the adhesive system 'self-etching' in the dentinal surfaces because there was a decline in the strength of adhesion in the groups conditioned with the laser. (author)

The tensile effect on crack formation in single crystal silicon irradiated by intense pulsed ion beam

Science.gov (United States)

Liang, Guoying; Shen, Jie; Zhang, Jie; Zhong, Haowen; Cui, Xiaojun; Yan, Sha; Zhang, Xiaofu; Yu, Xiao; Le, Xiaoyun

2017-10-01

Improving antifatigue performance of silicon substrate is very important for the development of semiconductor industry. The cracking behavior of silicon under intense pulsed ion beam irradiation was studied by numerical simulation in order to understand the mechanism of induced surface peeling observed by experimental means. Using molecular dynamics simulation based on Stillinger Weber potential, tensile effect on crack growth and propagation in single crystal silicon was investigated. Simulation results reveal that stress-strain curves of single crystal silicon at a constant strain rate can be divided into three stages, which are not similar to metal stress-strain curves; different tensile load velocities induce difference of single silicon crack formation speed; the layered stress results in crack formation in single crystal silicon. It is concluded that the crack growth and propagation is more sensitive to strain rate, tensile load velocity, stress distribution in single crystal silicon.

Mechanism of laser ablation for aqueous media irradiated under confined-stress conditions

International Nuclear Information System (INIS)

Oraevsky, A.A.; Jacques, S.L.; Tittel, F.K.

1995-01-01

Pulsed laser ablation of aqueous medium irradiated under conditions of temporal confinement of thermal stress is described. Time-resolved measurements of laser-induced transient stress waves with simultaneous imaging of ablation process by laser-flash photography were performed. Stress transients induced in aqueous solution of K 2 CrO 4 by ablative nanosecond laser pulses at 355 nm were studied by a broad-band lithium niobate acoustic transducer. Recoil momentum upon material ejection was measured from the temporal profiles of the acoustic transducer signal as a function of incident laser fluence. Cavitation bubbles produced in the irradiated volume during the tensile phase of thermoelastic stress were shown to drive material ejection at temperatures substantially below 100 degree C. Experimental data are evident that nanosecond-pulse laser ablation of aqueous media (when temporal stress-confinement conditions are satisfied) include the following two main stages of material ejection: (1) ejection of water microdroplets due to expansion and rupture of subsurface cavitation bubbles; (2) ejection of liquid streams with substantial volume upon collapse of initial crater and large cavitation bubbles in the depth of irradiated volume (after coalescence of smaller bubbles). copyright 1995 American Institute of Physics

Thermal, tensile and rheological properties of low density polyethylene (LDPE) processed irradiated by gamma-ray

International Nuclear Information System (INIS)

Ferreto, Helio F.R.; Oliveira, Ana C.F. de; Parra, Duclerc F.; Lugao, Ademar B.

2013-01-01

The aim of this paper is to investigate structural changes of low density polyethylene (LDPE) modified by ionizing radiation (gamma rays). The gamma radiation process for modification of commercial polymers is a widely applied technique to promote new physical-chemical and mechanical properties. Gamma irradiation originates free radicals which can induce chain scission or recombination, providing its annihilation, branching or crosslinking. The samples were prepare in hydraulic press in temperature 180 deg C after was irradiated with gamma source of 60 Co at doses of 5, 10, 20, 50 or 100 kGy at a dose rate of 5 kGy/h in inert atmosphere. The changes in molecular structure of LDPE, after gamma irradiations were evaluated using thermogravimetric analysis (TGA) and tensile machine and oscillatory rheology. The results showed the variations of the properties depending on the dose at each atmosphere. (author)

Induction Heating on Dynamic Tensile Tests in CEA Saclay

International Nuclear Information System (INIS)

Averty, X.; Yvon, P.; Duguay, C.; Pizzanelli, J. P.; Basini, V.

2001-01-01

The LCMI (Laboratory for characterization of irradiated materials), located in CEA from Saclay, is in charge of the mechanical tests on irradiated materials. The dynamic tensile testing machine, in a hot cell equipped with two remote handling, has been first improved in 1995, to fulfill the French safety programs on Reactivity Initiated Accident (RIA). One objective of this machine is to obtain mechanical property data on current Zircaloy cladding types needed to quality the cladding's response under RIA or LOCA transient loading and thermal conditions. For the RIA, this means testing at strain rates up to 5 s' and heating rates up to 200 degree centigree-s''-1, while for Loss of Coolant Accidents (LOCA) testing at strain rates of 10''-3 s''-1 and heating rates of 20 degree centigree s''-1 would be appropriate. The tensile samples are machined with a spark erosion machine, directly from pieces of cladding previously de fueled. Two kinds of samples can be machined in the cladding. Axial samples in order to test axial mechanical characteristics Ring samples in order to test transverse mechanical characteristics, more representative of RIA conditions. On one hand, the axial tensile tests were performed using the Joule effect, and heating rates up to about 500 degree centigree .s''-1 were obtained. This enabled us to perform the axial tests in a satisfactory manner. On the other hand, the tensile ring were first performed in a vertical furnace with a heating rate about 0.2 degree centigree.s''-1 and a thermal stability about 1 degree centigree. For temperatures above 480 degree centigree, the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently developed an induction heating system to reach heating rates high enough (200 degree centigree.s''-1) to prevent any significant annealing before performing the ring tensile tests. To apply a uniaxial tangential tension, two matching half

Tensile Bond Strengths of Two Adhesives on Irradiated and Nonirradiated Human Dentin

Directory of Open Access Journals (Sweden)

Cécile Bernard

2015-01-01

Full Text Available The aim of this study was to assess the effect of radiotherapy on bond efficiency of two different adhesive systems using tensile bond strength test. Twenty extracted teeth after radiotherapy and twenty nonirradiated extracted teeth were used. The irradiation was applied in vivo to a minimal dose of 50 Gy. The specimens of each group were randomly assigned to two subgroups to test two different adhesive systems. A three-step/etch-and-rinse adhesive system (Optibond FL and a two-steps/self-etch adhesive system (Optibond XTR were used. Composite buildups were performed with a nanohybrid composite (Herculite XTR. All specimens were submitted to thermocycling ageing (10000 cycles. The specimens were sectioned in 1 mm2 sticks. Microtensile bond strength tests were measured. Nonparametric statistical analyses were performed due to nonnormality of data. Optibond XTR on irradiated and nonirradiated teeth did not show any significant differences. However, Optibond FL bond strength was more effective on nonirradiated teeth than on irradiated teeth. Within the limitations of an in vitro study, it can be concluded that radiotherapy had a significant detrimental effect on bond strength to human dentin. However, it seems that adhesive choice could be adapted to the substrata. According to the present study, the two-steps/self-etch (Optibond XTR adhesive system tested could be more effective on irradiated dentin compared to three-steps/etch-and-rinse adhesive system (Optibond FL.

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Energy Technology Data Exchange (ETDEWEB)

Yano, Y., E-mail: yano.yasuhide@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T. [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Ukai, S.; Oono, N. [Materials Science and Engineering, Faculty of Engineering, Hokkaido University, N13, W-8, Kita-ku, Sapporo, Hokkaido, 060-8628 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hayashi, S. [Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Torimaru, T. [Nippon Nuclear Fuel Development Co., Ltd., 2163, Narita-cho, Oarai-machi, Ibaraki, 311-1313 (Japan)

2017-04-15

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Science.gov (United States)

Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

2017-04-01

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Tensile and fracture toughness properties of copper alloys and their HIP joints with austenitic stainless steel in unirradiated and neutron irradiated condition

International Nuclear Information System (INIS)

Taehtinen, S.; Pyykkoenen, M.; Singh, B.N.; Toft, P.

1998-03-01

The tensile strength and ductility of unirradiated CuAl25 IG0 and CuCrZr alloys decreased continuously with increasing temperature up to 350 deg C. Fracture toughness of unirradiated CuAl25 IG0 alloy decreased continuously with increasing temperature from 20 deg C to 350 deg C whereas the fracture toughness of unirradiated CuCrZr alloy remained almost constant at temperatures up to 100 deg C, was decreased significantly at 200 deg C and slightly increased at 350 deg C. Fracture toughness of HIP joints were lower than that of corresponding copper alloy and fracture path in HIP joint specimen was always within copper alloy side of the joint. Neutron irradiation to a dose level of 0.3 dpa resulted in hardening and reduction in uniform elongation to about 2-4% at 200 deg C in both copper alloys. At higher temperatures softening was observed and uniform elongation increased to about 5% and 16% for CuAl25 IG0 and CuCrZr alloys, respectively. Fracture toughness of CuAl25 IG0 alloy reduced markedly due to neutron irradiation in the temperature range from 20 deg C to 350 deg C. The fracture toughness of the irradiated CuCrZr alloy also decreased in the range from 20 deg C to 350 deg C, although it remained almost unaffected at temperatures below 200 deg C and decreased significantly at 350 deg C when compared with that of unirradiated CuCrZr alloy. (orig.)

Effects of ion irradiation on the mechanical properties of several polymers

International Nuclear Information System (INIS)

Sasuga, Tsuneo; Kawanishi, Shunichi; Nishi, Masanobu; Seguchi, Tadao

1991-01-01

The effects of high-energy ion irradiation on the tensile properties of polymers were studied under conditions in which ions should pass completely through the specimen and the results were compared with 2 MeV electron irradiation effects. Experiments were carried out on polymers having various constituents and molecular structures, i.e. eight aliphatic polymers and four aromatic polymers. In the aliphatic polymers studied there was scarcely any difference in the dose dependence of the tensile strength and ultimate elongation between proton and electron irradiation. In the aromatic polymers, however, the decrements in the tensile strength and ultimate elongation vs proton dose were less than those for electron irradiation. In heavy-ion irradiation, the radiation damage of PE (an aliphatic polymer) decreased with increase of LET, but no obvious LET effects were observed in PES (an aromatic polymer). (author)

Use of run statistics to validate tensile tests

International Nuclear Information System (INIS)

Eatherly, W.P.

1981-01-01

In tensile testing of irradiated graphites, it is difficult to assure alignment of sample and train for tensile measurements. By recording location of fractures, run (sequential) statistics can readily detect lack of randomness. The technique is based on partitioning binomial distributions

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

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Nowicki, L.; Smith, D.L. [Argonne National Lab., IL (United States)

1997-08-01

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

Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

Energy Technology Data Exchange (ETDEWEB)

Scibetta, M.; Chaouadi, R.; Puzzolante, J.L

1999-10-01

Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm{sup 2} (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain.

Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

International Nuclear Information System (INIS)

Scibetta, M.; Chaouadi, R.; Puzzolante, J.L.

1999-10-01

Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm 2 (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain

Tensile stress corrosion cracking of type 304 stainless steel irradiated to very high dose

Energy Technology Data Exchange (ETDEWEB)

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

2001-09-01

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20--100 displacement per atom or dpa) by the end of life. The data bases and mechanistic understanding of, the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-commotion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-11 reactor after irradiation to {approximately}50 dpa at {approximately}370 C. Slow-strain-rate tensile tests were conducted at 289 C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microcopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.

Simulation of tensile stress-strain properties of irradiated type 316 SS by heavily cold-worked material

International Nuclear Information System (INIS)

Muto, Yasushi; Jitsukawa, Shiro; Hishinuma, Akimichi

1995-07-01

Type 316 stainless steel is one of the most promising candidate materials to be used for the structural parts of plasma facing components in the nuclear fusion reactor. The neutron irradiation make the material brittle and reduces its uniform elongation to almost zero at heavy doses. In order to apply such a material of reduced ductility to structural components, the structural integrity should be examined and assured by the fracture mechanics. The procedure requires a formulated stress-strain relationship. However, the available irradiated tensile test data are very limited at present, so that the cold-worked material was used as a simulated material in this study. Property changes of 316 SS, that is, a reduction of uniform elongation and an enhancement of yield stress are seemingly very similar for both the irradiated 316 SS and the cold-worked one. The specimens made of annealed 316 SS, 20% (or 15%) cold worked one and 40% cold worked one were prepared. After the formulation of stress strain behavior, the equation for the cold-worked 316 SS was fitted to the data on irradiated material under the assumption that the yield stress is the same for both materials. In addition, the upper limit for the plastic strain was introduced using the data on the irradiated material. (author)

A combined experimental and FE analysis procedure to evaluate tensile behavior of zircaloy pressure tubes

International Nuclear Information System (INIS)

Samal, M.K.; Vaze, K.K.; Balakrishnan, K.S.; Anantharaman, S.

2012-01-01

Determination of transverse mechanical properties from the ring type of specimens directly machined from the nuclear reactor pressure tubes is not straightforward because of the presence of combined membrane as well as bending stresses arising in the loaded condition. In this work, we have performed ring-tensile tests on the un-irradiated ring tensile specimen using two split semi-cylindrical mandrels as the loading device. A 3-D finite element (FE) analysis was performed in order to determine the material true stress-strain curve by comparing experimental load-displacement data with those predicted by FE analysis. In order to validate the methodology, miniaturized tensile specimens were machined from these tubes and tested. It was observed that the stress-strain data as obtained from ring tensile specimen could describe the load displacement curve of the miniaturized flat tensile specimen very well. (author)

Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

Energy Technology Data Exchange (ETDEWEB)

Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Shiba, K.; Tanigawa, H.; Ando, M. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge National Laboratory, TN (United States); Stoller, R. [ORNL - Oak Ridge National Laboratory, Materials Science and Technology Div., Oak Ridge, AK TN (United States)

2007-07-01

Full text of publication follows: Post irradiation tensile tests on reduced activation ferritic/martensitic steel, F82H have been conducted over the past two decades using Japan Materials Testing Reactor (JMTR) of JAEA, and Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, USA, under Japan/US collaboration programs. According to these results, F82H does not demonstrate irradiation hardening above 673 K up to 60 dpa. The current study has been concentrated on hardening behavior at temperature around 573 K. A series of low temperature irradiation experiment has been conducted at the HFIR under the international collaborative research between JAEA/US-DOE. In this collaboration, the irradiation condition is precisely controlled by the well matured capsule designing and instrumentation. This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels compared with the irradiation properties database on F82H. Post irradiation tensile tests have been conducted on the F82H and its modified steels irradiated at 573 K and the dose level was up to 25 dpa. According to these results, irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated 0.2 % proof stress is less than 1 GPa at ambient temperature. The deterioration of total elongation was also saturated by 9 dpa irradiation. The ductility of some modified steels which showed larger total elongation than that of F82H before irradiation become the same level as that of standard F82H steel after irradiation, even though its magnitude of irradiation hardening is smaller than that of F82H. This suggests that the more ductile steel demonstrates the more ductility loss at this temperature, regardless to the hardening level. The difference in ductility loss behavior between various tensile specimens will be discussed as the ductility could depend on the specimen dimension. (authors)

A comparison of the iraddiated tensile properties of a high-manganese austenitic steel and type 316 stainless steel

International Nuclear Information System (INIS)

Klueh, R.L.; Grossbeck, M.L.

1984-01-01

The USSR steel EP-838 is a high-manganese, low-nickel steel that also has lower chromium and molybdenum than type 316 stainless steel. Tensile specimens of 20%-cold-worked EP-838 and type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at the coolant temperature (approx.=50 0 C). A displacement damage level of 5.2 dpa was reached for the EP-838 and up to 9.5 dpa for the type 316 stainless steel. Tensile tests at room temperature and 300 0 C on the two steels indicated that the irradiation led to increased strength and decreased ductility compared to the unirradiated steels. Although the 0.2% yield stress of the type 316 stainless steel in the unirradiated condition was greater than that for the EP-838, after irradiation there was essentially no difference between the strength or ductility of the two steels. The results indicate that the replacement of the majority of the nickel by manganese and a reduction of chromium and molybdenum in an austenitic stainless steel of composition near that for type 316 stainless steel has little effect on the irradiated and unirradiated tensile properties at low temperatures. (orig.)

Optimization of tensile method and specimen geometry in modified ring tensile test

International Nuclear Information System (INIS)

Kitano, Koji; Fuketa, Toyoshi; Sasajima, Hideo; Uetsuka, Hiroshi

2001-03-01

Several techniques in ring tensile test are proposed in order to evaluate mechanical properties of cladding under hoop loading condition caused by pellet/cladding mechanical interaction (PCMI). In the modified techniques, variety of tensile methods and specimen geometry are being proposed in order to limit deformation within the gauge section. However, the tensile method and the specimen geometry were not determined in the modified techniques. In the present study, we have investigated the tensile method and the specimen geometry through finite element method (FEM) analysis of specimen deformation and tensile test on specimens with various gauge section geometries. In using two-piece tensile tooling, the mechanical properties under hoop loading condition can be correctly evaluated when deformation part (gauge section) is put on the top of a half-mandrel, and friction between the specimen and the half-mandrel is reduced with Teflon tape. In addition, we have shown the optimum specimen geometry for PWR 17 by 17 type cladding. (author)

Ring ductility of irradiated Inconel 706 and Nimonic PE16

International Nuclear Information System (INIS)

Huang, F.H.; Fish, R.L.

1984-01-01

The tensile ductility of fast neutron-irradiated, precipitation-hardened alloys Inconel 706 and Nimonic PE16 has been observed to be very low for certain test conditions. Explanations for the low ductility behavior have been sought by examination of broken tensile specimens with microscopy and other similar techniques. A ring compression test provides a method of evaluating the ductility of irradiated cladding specimens. Unlike the conventional uniaxial tensile testing in which the tensile specimen is deformed uniformly, the ring specimen is subjected to localized bending where the crack is initiated. The ductility can be estimated through an analysis of the bending of a ring in terms of strain hardening. Ring sections from irradiated, solution-treated Inconel 706 and Nimonic PE16 were compressed in the diametral direction to provide load-deflection records over a wide range of irradiation and test temperatures. Results showed that ductility in both alloys decreased with increasing test temperatures. The poorest ductility was exhibited at different irradiation temperatures in the two alloys - near 550 0 C for PE16 and 460 to 520 0 C for Inconel 706. The ring ductility data indicate that the grain boundary strength is a major factor in controlling the ductility of the PE16 alloy

Thermal, tensile and rheological properties of high density polyethylene (HDPE) processed and irradiated by gamma-ray in different atmospheres

Energy Technology Data Exchange (ETDEWEB)

Ferreto, H. F. R., E-mail: hferreto@ipen.br, E-mail: ana-feitoza@yahoo.com.br; Oliveira, A. C. F., E-mail: hferreto@ipen.br, E-mail: ana-feitoza@yahoo.com.br; Parra, D. F., E-mail: dfparra@ipen.br, E-mail: ablugao@ipen.br; Lugão, A. B., E-mail: dfparra@ipen.br, E-mail: ablugao@ipen.br [Center of Chemistry and Environment, Institute of Energy and Nuclear Research - IPEN (Brazil); Gaia, R., E-mail: renan-gaia7@hotmail.com [Faculdades Oswaldo Cruz (Brazil)

2014-05-15

The aim of this paper is to investigate structural changes of high density polyethylene (HDPE) modified by ionizing radiation (gamma rays) in different atmospheres. The gamma radiation process for modification of commercial polymers is a widely applied technique to promote new physical-chemical and mechanical properties. Gamma irradiation originates free radicals which can induce chain scission or recombination, providing its annihilation, branching or crosslinking. This polymer was irradiated with gamma source of {sup 60}Co at doses of 5, 10, 20, 50 or 100 kGy at a dose rate of 5 kGy/h. The changes in molecular structure of HDPE, after gamma irradiations were evaluated using thermogravimetric analysis (TGA) and tensile machine and oscillatory rheology. The results showed the variations of the properties depending on the dose at each atmosphere.

Influence of thermal conditions on the tensile properties of basalt fiber reinforced polypropylene–clay nanocomposites

International Nuclear Information System (INIS)

Eslami-Farsani, Reza; Reza Khalili, S. Mohammad; Hedayatnasab, Ziba; Soleimani, Neda

2014-01-01

Highlights: • We studied tensile properties of basalt fiber/nanoclay-polypropylene (BF–PPCN). • Addition of nanoclay improves the yield strength and Young’s modulus of BF–PPCN. • The tensile properties of BF–PPCN are high at low temperature (−196 °C). - Abstract: In this paper, a comparative study on the tensile properties of clay reinforced polypropylene (PP) nanocomposites (PPCN) and chopped basalt fiber reinforced PP–clay nanocomposites (PPCN-B) is presented. PP matrix are filled with 1, 3 and 5 wt.% of nanoclays. The ultimate tensile strength, yield strength, Young’s modulus and toughness are measured at various temperature conditions. The thermal conditions are included the room temperature (RT), low temperature (LT) and high temperature (HT). The basal spacing of clay in the composites is measured by X-ray diffraction (XRD). Nanoscale morphology of the samples is observed by transmission electron microscopy (TEM). Addition of nanoclay improves the yield strength and Young’s modulus of PPCN and PPCN-B; however, it reduces the ultimate tensile strength. Furthermore, the addition of chopped basalt fibers to PPCN improves the Young’s modulus of the composites. The Young’s modulus and the yield strength of both PPCN and PPCN-B are significantly high at LT (−196 °C), descend at RT (25 °C) and then low at HT (120 °C)

Tensile properties of orthodontic elastomeric ligatures.

Science.gov (United States)

Ahrari, F; Jalaly, T; Zebarjad, M

2010-01-01

Tensile properties of elastomeric ligatures become important when efficiency of orthodontic appliances is considered. The aim of this study was to compare tensile strength, extension to tensile strength, toughness and modulus of elasticity of elastomeric ligatures in both the as--received condition and after 28 days of immersion in the simulated oral environment. Furthermore, the changes that occurred in tensile properties of each brand of ligatures after 28 days were evaluated. Experimental-laboratory based. Elastomeric ligatures were obtained from different companies and their tensile properties were measured using Zwick testing machine in both the as-received condition and after 28 days of immersion in the simulated oral environment. The data were analyzed using independent sample t-tests, analysis of variance and Tukey tests. After 28 days, all the ligatures experienced a significant decrease in tensile strength, extension to tensile strength and toughness ( P tensile properties of different brands of ligatures in both conditions ( P tensile properties of different brands of ligatures, which should be considered during selection of these products.

Study on irradiation condition in radiation vulcanization of natural rubber latex

International Nuclear Information System (INIS)

Zhu Nankang; Wang Chunlei; Makuuchi, K.; Yoshii, F.

1996-01-01

The effect of gamma rays irradiation dose rates on RVNRL preparation was studied using Malaysian latex added with O.2 phr of KOH and 5 phr n-BA. The results showed, to ensure the tensile strength of the latex film meet the requirement, when applying vulcanisation doses, Dv of 20 kGy 20 and 15 kGy, irradiation dose rates should not be greater than 0. 49 kGy/hr and 1. 6 kGy/hr respectively. Its was found that within the storage time of 20 days there was no change in the physical properties of the latex films

Effect of helium on tensile properties of vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Billone, M.C.; Smith, D.L. [Argonne National Lab., IL (United States)

1997-08-01

Tensile properties of V-4Cr-4Ti (Heat BL-47), 3Ti-1Si (BL-45), and V-5Ti (BL-46) alloys after irradiation in a conventional irradiation experiment and in the Dynamic Helium Charging Experiment (DHCE) were reported previously. This paper presents revised tensile properties of these alloys, with a focus on the effects of dynamically generated helium of ductility and work-hardening capability at <500{degrees}C. After conventional irradiation (negligible helium generation) at {approx}427{degrees}C, a 30-kg heat of V-4Cr-4Ti (BL-47) exhibited very low uniform elongation, manifesting a strong susceptibility to loss of work-hardening capability. In contrast, a 15-kg heat of V-3Ti-1Si (BL -45) exhibited relatively high uniform elongation ({approx}4%) during conventional irradiation at {approx}427{degrees}C, showing that the heat is resistant to loss of work-hardening capability.

Low dose irradiation effects on DIN 1.4948 mechanical properties

International Nuclear Information System (INIS)

Schaaf, B. van der; Vries, M.I. de

For the SNR 300 the licensing authorities require the determination of the lower boundaries of post-irradiation mechanical properties for DIN 1.4948 parent metal and welded joints. It has been established that with decreasing strain rate the post-irradiation tensile ductility decreases. A transition strain rate has been observed, above which there is no effect of irradiation on ductility. The transition strain rate increases with increasing temperature. Coarse grained heats show lower ultimate tensile strength above 800 K than fine grained heats. There is no significant effect of irradiation on load controlled high cycle fatigue with frequencies of 1 Hz or higher. In low cycle fatigue numbers of cycles to failure decrease with decreasing frequency. Increasing the test temperature reduces the number of cycles to failure even more. The frequency effect is more evident at 823 K. Parent metal has a better fatigue resistance than welded joints in unirradiated and irradiated condition. Creep strength is reduced by irradiation due to loss of ductility. It is shown that with increasing grain size the rupture strength decreases. The ductility of welded joints after irradiation is low, in some cases as low as 0.5% creep strain. After irradiation, tensile, creep and fatigue fracture surfaces show many more intergranular features than in the equivalent unirradiated condition. The promotion of intergranular fracture by irradiation and the consequent degradation of low strain rate mechanical properties is explained by the presence of helium on grain boundaries. Several measures to increase the helium content threshold can be taken, such as grain refinement, homogeneous boron distribution and promotion of helium bubble initiation. In cases where helium embrittlement is encountered, life reduction factors on unirradiated material properties must be applied

Irradiation of copper alloys in FFTF

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1984-01-01

Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

Various conditioning methods for root canals influencing the tensile strength of titanium posts

NARCIS (Netherlands)

Schmage, P.; Sohn, J.; Nergiz, I.; Ozcan, M.; Nergiz, [No Value

2004-01-01

Conditioning the root canal is frequently advised to achieve high post-retention when resin composite luting cements are used. However, Manufacturers’ instructions for this purpose differ widely from one another. The aim of this study was to compare the tensile bond strengths of passive, tapered,

Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

International Nuclear Information System (INIS)

Chung, H.M.; Strain, R.V.; Shack, W.J.

2001-01-01

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ∼50 dpa at ∼370 deg. C. Slow-strain-rate tensile tests were conducted at 289 degree sign C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at a low ECP, and this susceptibility led to a poor work-hardening capability and low ductility

Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

International Nuclear Information System (INIS)

Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

1989-08-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

Mechanical behavior of styrene grafted PVC films by electron beam irradiation

International Nuclear Information System (INIS)

Cardoso, Jessica R.; Moura, Eduardo; Somessari, Elisabeth S.R.; Silveira, Carlos G.; Paes, Helio A.; Souza, Carlos A.; Manzoli, Jose E.; Geraldo, Aurea B.C.

2011-01-01

The polyvinyl chloride (PVC) is a technological and low cost polymer, however it presents high sensitivity to high energy irradiation because of the weakness of carbon-chloride bond face to carbon-carbon and carbon-hydrogen bonds. Grafting is a type of co-polymerization process that can allow it an increase of mechanical characteristics. The aim of this work is to evaluate the mechanical properties of styrene grafted PVC by electron beam irradiation using mutual and pre-irradiation methods to verify the mechanical resistance changes of obtained product whether grafting process is applied from non-irradiated or from pre-irradiated substrates. The irradiation procedures were performed in atmosphere air or inert atmosphere and the irradiation conditions comprised doses from 10 kGy to 100 kGy and dose rates of 2.2 kGy/s and 22.4 kGy/s. The styrene grafted samples were analyzed by gravimetry to determinate the grafting yield; the final values have been averaged from a series of three measurements. The Mid-A TR-FTIR was the spectrophotometer technique used for qualitative/semi-quantitative analysis of grafted samples. The Young's module and tensile strength of pre-irradiated and grafted PVC samples at both methods were measured at a Lloyd LXR tensile tester at a cross-head speed of 10.00 mm/min. We observed the decrease of Young's module and tensile strength with the increase of absorbed dose at pre-irradiated PVC samples. These mechanical parameters results are discussed. (author)

An Electron Microscope Study of the Thermal Neutron Induced Loss in High Temperature Tensile Ductility of Nb Stabilized Austenitic Steels

International Nuclear Information System (INIS)

Roy, R.B.

1965-04-01

Irradiated ∼3 x 10 19 n/cm 2 (thermal), 18 n/cm 2 (> 1 MeV) at 40 deg C and the corresponding unirradiated control tensile specimens of a 20 % Cr, 25 % Ni, Nb stabilized steel tested at 650 deg C, 750 deg C and 800 deg C have been examined by transmission electron microscopy. The results indicate that the irradiation induced embrittlement of the tensile specimens at elevated temperatures is preceded by the formation of fine precipitates within the grains. These precipitates may restrict the deformation within the grains such that the stresses are concentrated at the grain boundaries thereby leading to premature failure. It is suggested that the main effect of the irradiation is to promote conditions necessary for the formation of these precipitates, namely, super saturation and fresh nucleation sites within the matrix through the energetic emission of He and Li atoms from boron as an impurity

Irradiation effects on tensile ductility and dynamic toughness of ferritic-martensitic 7-12 Cr steels

International Nuclear Information System (INIS)

Preininger, D.

2006-01-01

The superimposed effect of irradiation-induced hardening by small defects (clusters, dislocation loops) and chromium-rich - precipitate formations on tensile ductility and Charpy-impact behaviour of various ferritic-martensitic (7-13)CrWVTa(Ti)-RAFM steels have been examined by micro-mechanical deformation and ductile/dynamic fracture models. Analytical relations have been deduced describing irradiation-induced changes of uniform ductility and fracture strain as well as ductile-to-brittle transition temperature DBTT and ductile upper shelf energy USE observed from impact tests. The models apply work-hardening with competitive action of relevant dislocation multiplication and annihilation reactions. The impact model takes into account stress intensity with local plasticity and fracture within the damage zone of main crack. Especially, the influences of radiation-induced changes in ductile and dynamic fracture stresses have been considered together with effects from strain rate sensitivity of strength, precipitate morphology as mean size dp and volume fraction fv as well as deformation temperature and strain rate. For these, particularly the correlation between tensile ductility and impact properties have been examined. Strengthening by clusters and loops generally reduces uniform ductility, and more stronger fracture strain as well as ductile upper shelf energy USE and additionally increases DBTT for constant fracture stresses. A superimposed precipitation hardening by formation of 3-6 nm, f v 6 nm, which clear above the sharable limit of coherent precipitates increases with increasing fraction fv and but strongly reduces with increasing matrix strength due to full martensitic structure, higher C, N alloying contents and pronounced hardening by irradiation-induced cluster and loop formations. A combined increase of fracture stresses due to irradiation-induced changes of the grain boundary structure diminishes the strength-induced increase in DBTT and more stronger

Aversive conditioning in prenatally gamma-irradiated rats

International Nuclear Information System (INIS)

Tamaki, Yoshitaka; Hoshino, Kiyoshi; Kameyama, Yoshiro

1987-01-01

To examine how intrauterine exposure to gamma rays would exert on four kinds of aversive conditioning, rat fetuses were irradiated with 0.27, 0.48, or 1.46 Gy at Day 15 post conception. When ordinary avoidance conditioning was given to the groups with 0.27 and 0.48 Gy, there was no significant difference between the irradiated groups and the control group in the rate of positive avoidance response. Nor was this different in the irradiated groups and the control group, when the rate of baseline response was examined in avoidance conditioning. In positive avoidance conditioning to two kinds of anticipatory electric stimuli, the acquisition of avoidance was significantly inferior in all irradiated groups to that in the control group. When giving succesive discrimination learning, the group with 1.46 Gy tended to have higher rate of positive avoidance response and remarkably lower rate of passive avoidance response than the control group. (Namekawa, K.)

Impact of some environmental conditions on the tensile, creep-recovery, relaxation, melting and crystallinity behaviour of UHMWPE-GUR 410-medical grade

International Nuclear Information System (INIS)

Mourad, A.-H.I.; Fouad, H.; Elleithy, Rabeh

2009-01-01

The present work was undertaken to examine the effect of some environmental media (sodium hydroxide NaOH solution, water, ice, UV irradiation dose and pre-heat treatment) on the mechanical (quasi-static tensile creep-recovery and relaxation) and physical/thermal (melting and crystallinity) behaviour of the ultra high molecular weight polyethylene (UHMWPE-GUR 410-medical grade), that has several biomedical and engineering applications. The results show changes in the mechanical properties due to these environmental effects. The pre-heat treatment has significantly enhanced the tensile properties compared to virgin specimens' properties. Improvement due to pre-heat treatment at 100 o C is more than that at 50 o C. Specimens' storing in ice, NaOH and water has not affected significantly the tensile properties. All properties except fracture strain have enhanced due to specimens exposure to UV irradiation. The differential scanning calorimetry results indicate that environmental media have not any noticeable effects on the melting temperature. However, a significant increase in the degree of crystallinity was observed for all specimens versus that for virgin specimens. The creep and permanent strains of the tested virgin material increase with temperature and lineally increase with applied load. The specimens' exposure to environmental media has improved the creep resistance and the permanent creep strain when compared with that for virgin ones. Remarkable increase was observed in the initial relaxation and residual stress of the exposed specimens against that for virgin specimens.

Study of tensile test behavior of austenitic stainless steel type 347 seamless thin-walled tubes in cold worked condition

Energy Technology Data Exchange (ETDEWEB)

Terui, Clarice, E-mail: clarice.terui@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CINA/CTMSP), Iperó, SP (Brazil). Centro Industrial Nuclear da Marinha; Lima, Nelson B. de, E-mail: nblima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

These austenitic stainless steel type 347 seamless thin-walled tubes are potential candidates to be used in fuel elements of nuclear power plants (as PWR - Pressurized Water Reactor). So, their metallurgical condition and mechanical properties, as the tensile strength and yield strength, normally are very restrict in demanding project and design requirements. Several full size tensile tests at room temperature and high temperature (315 deg C) were performed in these seamless tubes in cold-worked condition. The results of specified tensile and yield strengths were achieved but the elongation of the tube, in the geometry of the component, could not be measured at high temperature due to unconventional mode of rupture (helical mode without separation of parts). The average value of elongation was obtained from stress-strain curves of hot tensile tests and was around 5%. The results obtained in this research show that this behavior of the full size tensile test samples of thin-walled tube (wall thickness less than 0.5 mm) in high temperature (315°C) is due to the combination of the manufacturing process, the material (crystallographic structure and chemical composition) and the final geometry of the component. In other words, the strong crystallographic texture of material induced by tube drawing process in addition with the geometry of the component are responsible for the behavior in hot uniaxial tensile tests. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Pawel-Robertson, J.E.; Rowcliffe, A.F.; Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)] [and others

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.

Post-irradiation creep properties of four plates and two forgings DIN 1.4948 steel from the SNR-300 permanent primary structures

International Nuclear Information System (INIS)

Schaaf, B. van der.

1987-01-01

The safety authorities, involved in the licensing procedure of the SNR-300, have required the determination of the irradiation effect on the heat-to-heat variation of tensile and creep properties of Werkst. No. DIN 1.4948 austenitic stainless steel. These data are lacking in the present codes and they are necessary for the design and safety considerations of the permanent structures. Results are presented of about 200 tests on irradiated and unirradiated material of 6 heats used in the production of the SNR-300 permanent structures. After irradiation in the HFR-Petten to neutron fluences relevant for the SNR-300 service conditions post-irradiation tensile and creep tests (up to 10,000 hrs rupture time) were performed in the temperature range 723 K to 923 K. All heats are embrittled by irradiation resulting in reduction of rupture times, creep strength and ultimate tensile strength. The considerable reduction is attributed to helium enhanced intergranular creep crack growth, which reduces the ductility and strength, but does not affect the creep rate. The variation of tensile and creep properties is large and independent of irradiation. The minimum derived creep strength in irradiated condition drops below the values expected in the ASME Code and VdTuV Blatt. In design and safety analyses the irradiation effect on creep properties must be accounted for with an appropriate reduction factor. The predictions given, have to be verified with long-term creep tests and parts of the SNR surveillance programme. 172 figs.; 17 refs.; 58 tables

Nuclear irradiation parameters of beryllium under fusion, fission and IFMIF irradiation conditions

International Nuclear Information System (INIS)

Fischer, U.; Chen, Y.; Leichtle, D.; Simakov, S.; Moeslang, A.; Vladimirov, P.

2004-01-01

A computational analysis is presented of the nuclear irradiation parameters for Beryllium under irradiation in typical neutron environments of fission and fusion reactors, and of the presently designed intense fusion neutron source IFMIF. The analysis shows that dpa and Tritium production rates at fusion relevant levels can be achieved with existing high flux fission reactors while the achievable Helium production is too low. The resulting He-Tritium and He/dpa ratios do not meet typical fusion irradiation conditions. Irradiation simulations in the medium flux test modules of the IFMIF neutron source facility were shown to be more suitable to match fusion typical irradiation conditions. To achieve sufficiently high production rates it is suggested to remove the creep-fatigue testing machine together with the W spectra shifter plate and move the tritium release module upstream towards the high flux test module. (author)

An Electron Microscope Study of the Thermal Neutron Induced Loss in High Temperature Tensile Ductility of Nb Stabilized Austenitic Steels

Energy Technology Data Exchange (ETDEWEB)

Roy, R B

1965-04-15

Irradiated {approx}3 x 10{sup 19} n/cm{sup 2} (thermal), <3 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 40 deg C and the corresponding unirradiated control tensile specimens of a 20 % Cr, 25 % Ni, Nb stabilized steel tested at 650 deg C, 750 deg C and 800 deg C have been examined by transmission electron microscopy. The results indicate that the irradiation induced embrittlement of the tensile specimens at elevated temperatures is preceded by the formation of fine precipitates within the grains. These precipitates may restrict the deformation within the grains such that the stresses are concentrated at the grain boundaries thereby leading to premature failure. It is suggested that the main effect of the irradiation is to promote conditions necessary for the formation of these precipitates, namely, super saturation and fresh nucleation sites within the matrix through the energetic emission of He and Li atoms from boron as an impurity.

Mechanical behavior of styrene grafted PVC films by electron beam irradiation

Energy Technology Data Exchange (ETDEWEB)

Cardoso, Jessica R.; Moura, Eduardo; Somessari, Elisabeth S.R.; Silveira, Carlos G.; Paes, Helio A.; Souza, Carlos A.; Manzoli, Jose E.; Geraldo, Aurea B.C., E-mail: ageraldo@ipen.br, E-mail: jmanzoli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The polyvinyl chloride (PVC) is a technological and low cost polymer, however it presents high sensitivity to high energy irradiation because of the weakness of carbon-chloride bond face to carbon-carbon and carbon-hydrogen bonds. Grafting is a type of co-polymerization process that can allow it an increase of mechanical characteristics. The aim of this work is to evaluate the mechanical properties of styrene grafted PVC by electron beam irradiation using mutual and pre-irradiation methods to verify the mechanical resistance changes of obtained product whether grafting process is applied from non-irradiated or from pre-irradiated substrates. The irradiation procedures were performed in atmosphere air or inert atmosphere and the irradiation conditions comprised doses from 10 kGy to 100 kGy and dose rates of 2.2 kGy/s and 22.4 kGy/s. The styrene grafted samples were analyzed by gravimetry to determinate the grafting yield; the final values have been averaged from a series of three measurements. The Mid-A TR-FTIR was the spectrophotometer technique used for qualitative/semi-quantitative analysis of grafted samples. The Young's module and tensile strength of pre-irradiated and grafted PVC samples at both methods were measured at a Lloyd LXR tensile tester at a cross-head speed of 10.00 mm/min. We observed the decrease of Young's module and tensile strength with the increase of absorbed dose at pre-irradiated PVC samples. These mechanical parameters results are discussed. (author)

Design and fabrication of hafnium tube to control the power of the irradiation test fuel in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Lee, C. B.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H

2003-05-01

For the irradiation test at HANARO, non-instrumentation capsule was manufactured and hafnium tube was used to control LHGR of HANARO. Hafnium tube can control the irradiation condition of HANARO similar to that of commercial reactor. Hafnium tube thickness was determined by the LHGR calculated at OR-4 irradiation hole to be installed the non-instrumented capsule. To fabricate the hafnium tube with hafnium plate, the fabrication method was determined by using the hafnium mechanical properties. And the tensile strength of hafnium was confirmed by tensile test. This report is confirmed the LHGR control at the OR-4 and the Hafnium fabrication for in used which the AFPCAP non-instrumented irradiation capsule.

Spherical Nanoindentation Stress-Strain Measurements of BOR-60 14YWT-NFA1 Irradiated Tubes

Energy Technology Data Exchange (ETDEWEB)

Weaver, Jordan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carvajal Nunez, Ursula [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Krumwiede, David [Univ. of California, Berkeley, CA (United States); Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hosemann, Peter [Univ. of California, Berkeley, CA (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mara, Nathan Allan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

Spherical nanoindentation stress-strain protocols were applied to characterize unirradiated and fast neutron irradiated nanostructured ferritic alloy (NFA) 14YWT and compared against Berkovich nanohardness and available tensile data. The predicted uniaxial yield strength from spherical, 100 and 5 micron radii, indentation yield strength measurements was 1100-1400 MPa which compares well with the predictions from Berkovich nanohardness, 1200 MPa, and available tensile data, ~1100 MPa. However, spherical indentation measurements predict an increase in the uniaxial yield strength of ~1 GPa while Berkovich nanohardness measurements predict an increase of only ~250 MPa. No tensile data exists on the irradiated condition. It is believed the difference in the predicted uniaxial yield strength between spherical and Berkovich nanoindentation are due to a low number of tests on the irradiated sample combined with the significant heterogeneity in the microstructure, the differences in sensitivity to sample preparation on the irradiated sample between the two indentation protocols , and/or in how strain localizes under the indenter with the possibility of dislocation channeling under Berkovich hardness indents leading to strain softening. Nanoindentation capabilities to test neutron irradiated samples in a radiological area were realized.

Microstructure in HIP-bonded F82H steel and its mechanical properties after irradiation

International Nuclear Information System (INIS)

Furuya, K.; Wakai, E.

2006-01-01

A first primary blanket structure is composed of the low-activation steel, e.g. F82H, and is fabricated by using a solid hot isostatic pressing (HIP) bonding method. A partial mock-up of such a blanket structure was successfully fabricated. The tensile specimen including HIP-bonded region possessed a sufficient strength and elongation under a non-irradiated condition as reported in our previous studies. In this study, the microstructures of HIP interface before irradiation were observed by a TEM, and the effects of irradiation on mechanical properties of the HIP-bonded region were also examined. TEM observation and elemental analysis of the HIP-bonded region before the irradiation were performed by using a FE-TEM of HF-2000 equipped with EDX spectroscopy. Tensile specimens (type SS-3) were prepared from a HIP-bonded region and a plate region of the mock-up block. Neutron irradiation was performed up to about 1.9 dpa at about 523 K in JMTR. After the irradiation, tensile test was performed at temperatures of 295 and 523 K. After the tensile test, OM observation at the rupture region and SEM observation at the fracture surface were conducted, respectively. TEM observation and analytical results revealed that the HIP interface possessed many precipitates, and enriched peak spectrum of chromium was detected from the precipitates. In addition, aspect of the spectrum was qualitatively equivalent to that of M23C6 in grain boundaries of F82H steel. In result, the HIP boundary has many M23C6 which were generally seen in grain boundaries of F82H steel, and it can be mentioned that the HIP interface is, in this sense, a new grain boundary. Obvious HIP boundary was seen at rupture region of tensile specimens sampled from the HIP-bonded region, by the macroscopic observation. It means that rupture do not occur in the HIP interface. In result, it can be mentioned that bondability of the HIP interfaces is kept under the irradiation and testing conditions. The strength and

Evaluation of microtensile and tensile bond strength tests ...

African Journals Online (AJOL)

2015-11-03

Nov 3, 2015 ... Bond strength tests and Er,Cr:YSGG laser frequency. 586 ... power, 90% air pressure, 75% water pressure, 45 s irradiation ..... geometry on the measurement of the tensile bond strength to dentin. J Dent ... Bur‑cut enamel and.

Characterization of Irradiated and Non-Irradiated Rubber from Automotive Scrap Tires

Science.gov (United States)

Souza, Clécia Moura; Silva, Leonardo G.

The aim of this work was to characterize the samples of irradiated and non-irradiated rubber from automotive scrap tires. Rubber samples from scrap tires were irradiated at irradiation doses of 200, 400 and 600kGy in an electron beam accelerator. Subsequently, both the irradiated and non-irradiated samples were characterized by thermogravimetry (TG), differential scanning calorimetry (DSC), tensile strength mechanical test, and Fourier transform infrared (FTIR) spectrophotometry.

How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

CERN Document Server

Daum, E

2000-01-01

The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the sup 6 Li(n,t) sup 4 He channel as it occurs in a DEMO breeding blanket.

The ARBOR irradiation project

International Nuclear Information System (INIS)

Petersen, C.; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J.

2002-01-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10 15 n/cm 2 s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten

Design and use of nonstandard tensile specimens for irradiated materials testing

International Nuclear Information System (INIS)

Panayotou, N.F.

1984-10-01

Miniature, nonstandard, tensile-type specimens have been developed for use in radiation effects experiments at high energy neutron sources where the useful radiation volume is as small as a few cubic centimeters. The end result of our development is a sheet-type specimen, 12.7 mm long with a 5.1 mm long, 1.0 mm wide gage section, which is typically fabricated from 0.25 mm thick sheet stock by a punching technique. Despite this miniature geometry, it has been determined that the data obtained using these miniature specimens are in good agreement with data obtained using much larger specimens. This finding indicates that miniature tensile specimen data may by used for engineering design purposes. Furthermore, it is clear that miniature tensile specimen technology is applicable to fields other than the study of radiation effects. This paper describes the miniature specimen technology which was developed and compares the data obtained from these miniature specimens to data obtained from much larger specimens. 9 figures

Low cycle fatigue of irradiated LMFBR materials

International Nuclear Information System (INIS)

Blackburn, L.D.

1976-01-01

A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

Effects of reclaimed asphalt pavement on indirect tensile strength test of conditioned foamed asphalt mix

International Nuclear Information System (INIS)

Katman, Herda Yati; Norhisham, Shuhairy; Ismail, Norlela; Ibrahim, Mohd Rasdan; Matori, Mohd Yazip

2013-01-01

This paper presents the results of Indirect Tensile Strength (ITS) Test for samples prepared with reclaimed asphalt pavement (RAP). Samples were conditioned in water at 25°C for 24 hours prior to testing. Results show that recycled aggregate from reclaimed asphalt pavement performs as well as virgin aggregate.

Post-irradiation mechanical tests on F82H EB and TIG welds

International Nuclear Information System (INIS)

Rensman, J.; Osch, E.V. van; Horsten, M.G.; D'Hulst, D.S.

2000-01-01

The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2-3 dpa at a temperature of 300 deg. C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250 deg. C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition

Continuum model of tensile fracture of metal melts and its application to a problem of high-current electron irradiation of metals

International Nuclear Information System (INIS)

Mayer, Alexander E.; Mayer, Polina N.

2015-01-01

A continuum model of the metal melt fracture is formulated on the basis of the continuum mechanics and theory of metastable liquid. A character of temperature and strain rate dependences of the tensile strength that is predicted by the continuum model is verified, and parameters of the model are fitted with the use of the results of the molecular dynamics simulations for ultra-high strain rates (≥1–10/ns). A comparison with experimental data from literature is also presented for Al and Ni melts. Using the continuum model, the dynamic tensile strength of initially uniform melts of Al, Cu, Ni, Fe, Ti, and Pb within a wide range of strain rates (from 1–10/ms to 100/ns) and temperatures (from melting temperature up to 70–80% of critical temperature) is calculated. The model is applied to numerical investigation of a problem of the high-current electron irradiation of Al, Cu, and Fe targets

The ARBOR irradiation project

Energy Technology Data Exchange (ETDEWEB)

Petersen, C. E-mail: claus.petersen@imf.fzk.de; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J

2002-12-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10{sup 15} n/cm{sup 2} s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten.

A Study on Mechanical behavior of Tensile Specimen Fabricated by Laser Cutting

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The mechanical testing data are required for the assessment of dry storage of the spent nuclear fuel. Laser cutting system could be useful tools for material processing such as cutting in radioactive environment due to non-contact nature, ease in handling and the laser cutting process is most advantageous, offering the narrow kerf width and heat affected zone by using small beam spot diameter. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of the unirradiated cladding with and without oxide layer on the specimens. In the present study, the dimensional measurement and tensile test were conducted to investigate the mechanical behavior of the axial tensile test specimens depending on the material processing methods in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser cutting system was used to fabricate the tensile test specimens, and the mechanical behavior was investigated using the dimensional measurement and tensile test. It was shown that the laser beam machining could be a useful tool to fabricate the specimens and this technique will be developed for the fabrication of various types of irradiated specimens in a hotcell.

A Study on Mechanical behavior of Tensile Specimen Fabricated by Laser Cutting

International Nuclear Information System (INIS)

Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y.

2016-01-01

The mechanical testing data are required for the assessment of dry storage of the spent nuclear fuel. Laser cutting system could be useful tools for material processing such as cutting in radioactive environment due to non-contact nature, ease in handling and the laser cutting process is most advantageous, offering the narrow kerf width and heat affected zone by using small beam spot diameter. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of the unirradiated cladding with and without oxide layer on the specimens. In the present study, the dimensional measurement and tensile test were conducted to investigate the mechanical behavior of the axial tensile test specimens depending on the material processing methods in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser cutting system was used to fabricate the tensile test specimens, and the mechanical behavior was investigated using the dimensional measurement and tensile test. It was shown that the laser beam machining could be a useful tool to fabricate the specimens and this technique will be developed for the fabrication of various types of irradiated specimens in a hotcell

Graft Copolymerization of Methyl Methacrylate Monomer onto Starch and Natural Rubber Latex Initiated by Gamma Irradiation

Directory of Open Access Journals (Sweden)

S. Iskandar

2011-04-01

Full Text Available To obtain the degradable plastic, the graft copolymerization of methyl methacrylate onto starch and natural rubber latex was conducted by a simultaneous irradiation technique. Gamma-ray from cobalt-60 source was used as the initiator. The grafted copolymer of starch-polymethyl methacrylate and the grafted copolymer of natural rubber-polymethyl methacrylate were mixed in the blender, and dried it in the oven. The dried grafted copolymer mixture was then molded using hydraulic press machine. The effect of irradiation dose, composition of the grafted copolymer mixture, film forming condition and recycle effect was evaluated. The parameters observed were tensile strength, gel fraction and soil burial degradability of grafted copolymer mixture. It was found that the tensile strength of grafted copolymer mixture increased by -ray irradiation. Increasing of the grafted copolymer of natural rubber-polymethyl methacrylate content, the gel fraction and tensile strength of the grafted copolymer mixture increased. The tensile strength of the grafted copolymer mixture was increased from 18 MPa to 23 MPa after recycled (film forming reprocessed 3 times. The grafted copolymer mixture was degraded completely after soil buried for 6 months

Tensile and fracture properties of EBR-II-irradiated V-15Cr-5Ti containing helium

International Nuclear Information System (INIS)

Grossbeck, M.L.; Horak, J.A.

1986-01-01

The alloy V-15Cr-5Ti was cyclotron-implanted with 80 appM He and subsequently irradiated in the Experimental Breeder Reactor (EBR-II) to 30 dpa. The same alloy was also irradiated in the 10, 20, and 30% cold-worked conditions. Irradiation temperatures ranged from 400 to 700 0 C. No significant effects of helium on mechanical properties were found in this temperature range although the neutron irradiation shifted the temperature of transition from cleavage to ductile fracture to about 625 0 C. Ten percent cold work was found to have a beneficial effect in reducing the tendency for cleavage fracture following irradiation, but high levels (20%) were observed to reduce ductility. Still higher levels (30%) improved ductility by inducing recovery during the elevated-temperature irradiation. Swelling was found to be negligible, but precipitates - titanium oxides or carbonitrides - contained substantial cavities

Elastic-plastic analysis of the SS-3 tensile specimen

International Nuclear Information System (INIS)

Majumdar, S.

1998-01-01

Tensile tests of most irradiated specimens of vanadium alloys are conducted using the miniature SS-3 specimen which is not ASTM approved. Detailed elastic-plastic finite element analysis of the specimen was conducted to show that, as long as the ultimate to yield strength ratio is less than or equal to 1.25 (which is satisfied by many irradiated materials), the stress-plastic strain curve obtained by using such a specimen is representative of the true material behavior

Influence of gamma-radiation on tensile strength properties of polytetrafluoroethylene (PTFE)

CERN Document Server

Gafurov, U G; Nemkova, N

2002-01-01

The tensile strength properties of polytetrafluoroethylene are studied at modification doses of gamma-irradiation. The main molecular process of polymer destruction is found to be the thermostimulated slippage of molecular chains. (author)

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

The postirradiation tensile properties and microstructure of several vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1988-01-01

Tensile specimens of V-15Cr-5Ti, Vanstar-7, V-3Ti-1si, and V-20Ti were irradiated at 420/degrees/C in FFTF-MOTA to a damage level of 82 dpa. Helium was preimplanted to levels up to 480 appm in selected specimens using a modified tritium trick. Irradiation hardening was the dominant effect influencing the postirradiation tensile properties, and it markedly increased the yield strength and reduced the total elogation. The V-15Cr-5Ti alloy was very sensitive to helium embrittlement, but Vanstar-7 and V-3Ti-1Si were only slightly affected. Without helium, negligible swelling (<1%) were measured in V-3Ti-1Si and V-20Ti. Preimplanted helium increased swelling in V-3Ti-1Si by increasing cavity nucleation. 11 refs., 11 figs., 3 tabs

Tensile and fracture properties of EBR-II-irradiated V-15Cr-5Ti containing helium

Energy Technology Data Exchange (ETDEWEB)

Grossbeck, M.L.; Horak, J.A.

1986-01-01

The alloy V-15Cr-5Ti was cyclotron-implanted with 80 appM He and subsequently irradiated in the Experimental Breeder Reactor (EBR-II) to 30 dpa. The same alloy was also irradiated in the 10, 20, and 30% cold-worked conditions. Irradiation temperatures ranged from 400 to 700/sup 0/C. No significant effects of helium on mechanical properties were found in this temperature range although the neutron irradiation shifted the temperature of transition from cleavage to ductile fracture to about 625/sup 0/C. Ten percent cold work was found to have a beneficial effect in reducing the tendency for cleavage fracture following irradiation, but high levels (20%) were observed to reduce ductility. Still higher levels (30%) improved ductility by inducing recovery during the elevated-temperature irradiation. Swelling was found to be negligible, but precipitates - titanium oxides or carbonitrides - contained substantial cavities.

Electrical in situ and post-irradiation properties of ceramics relevant to fusion irradiation conditions

International Nuclear Information System (INIS)

Shikama, Tatsuo; Zinkle, S.J.

2002-01-01

Electrical properties of ceramic candidate materials for the next-generation nuclear fusion devices under relevant irradiation conditions are reviewed. A main focal point is placed on the degradation behavior of the electrical insulating ability during and after irradiation. Several important radiation induced effects play important roles: radiation induced conductivity, thermally stimulated electrical conductivity, radiation induced electrical charge separation, and radiation induced electromotive force. These phenomena will interact with each other under fusion relevant irradiation conditions. The design of electrical components for the next-generation fusion devices should take into account these complicated interactions among the radiation induced phenomena

Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

Energy Technology Data Exchange (ETDEWEB)

Yamashita, Shinichiro, E-mail: yamashita.shinichiro@jaea.go.jp; Yano, Yasuhide; Ohtsuka, Satoshi; Yoshitake, Tsunemitsu; Kaito, Takeji; Koyama, Shin-ichi; Tanaka, Kenya

2013-11-15

Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS{sub F}/M, 12Cr–ODS{sub F}) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS{sub F}/M steel whereas no distinct temperature dependence for 12Cr–ODS{sub F} steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10{sup 26} n/m{sup 2} (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

Study on creep-fatigue life of irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji

2001-01-01

The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2 dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. The creep damage of both unirradiated and irradiated specimens was underestimated by the time fraction rule or the ductility exhaustion rule. The creep damage calculated by the time fraction rule or the ductility exhaustion rule increased by the irradiation. The predictions derived from the linear damage rule are unsafe as compared with the experimental fatigue lives. (author)

Irradiation hardening of Mod.9Cr-1Mo steel

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

2009-01-01

An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

European Fusion Programme. ITER task T23: Beryllium characterisation. Progress report. Tensile tests on neutron irradiated and reference beryllium

International Nuclear Information System (INIS)

Moons, F.

1996-02-01

As part of the European Technology Fusion Programme, the irradiation embrittlement characteristics of the more ductile and isotopic grades of beryllium manufactured by Brush Wellman has been investigated using modern powder production and consolidation techniques . This study was initiated in support of the development and evaluation of beryllium as a neutron multiplier for the solid breeder blanket design concepts proposed for a DEMO fusion power reactor. Four different species of beryllium: S-200 F (vacuum hot pressed, 1.2 wt% BeO), S-200FH (hot isostatic pressed, 0.9 wt% BeO), S-65 (vacuum hot pressed, 0.6 wt% BeO), S-65H (hot isostatic pressed, 0.5 wt% BeO) have been compared. Three batches of the beryllium have been investigated, a neutron batch, a thermal control batch and a reference batch. Neutron irradiation has been performed at temperatures between 175 and 605 degrees Celsius up to a neutron fluence of 2.1 10 25 n.m -2 (E> 1 MeV) or 750 appm He. The results of the tensile tests are summarized

Coolant compatibility studies. The effect of irradiation on tensile properties and stress corrosion cracking sensitivity of martensitic steels. MANET 4 - complementary studies

International Nuclear Information System (INIS)

Nystrand, A.C.

1994-02-01

Tensile and stress corrosion cracking tests have been carried out on MANET-type (1.4914 and FV448) and reduced activation (LA12TaLC) high-chromium martensitic steels. The materials had previously been exposed up to 5000 h at ∼275 degrees C in the core, above the core and remote from the core of a high pressure water loop in the Studsvik R2 reactor. After the mechanical testing the materials were examined visually and metallographically. The steel samples exposed in the core section showed large increases in tensile yield strengths when tested at 250 degrees C. However, the magnitude of the radiation hardening was considerably smaller in the reduced activation steel compared to the commercial steels; this observation is consistent with published data on other high-chromium martensitic steels and is associated with the lower chromium content of the LA12TaLC steel (8.9%) compared with those of the commercial steels (10.6 and 11.3%). Irradiation assisted stress corrosion cracking (IASCC) was not detected in any of the stressed steel samples after autoclave testing for times up to 1500 h at 250 degrees C in air-saturated high purity water. This apparent resistance to IASCC may be due to the high chromium martensitic steels not being sensitized by the irradiation in a comparable manner to that shown by the austenitic steels. However, additional studies are required to clarify some of the existing uncertainties with respect to IASCC of these martensitic steels

Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 deg. C

International Nuclear Information System (INIS)

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

1997-02-01

Screening experiments were carried out to investigate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) cooper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing, and bonding thermal treatment followed by re-ageing and the reactor bakeout treatment at 350 deg. C for 100 h. Tensile specimens of CuAl-25 were given the heat treatment corresponding to the bonding thermal cycle. A number of heat treated specimens were neutron irradiated at 350 deg. C to a dose level of ∼ 0.3 dpa in the DR-3 reactor at Risoe. Both unirradiated and irradiated specimens with various heat treatments were tensile tested at 350 deg. C. The microstructure and electrical resistivity of these specimens were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure of the irradiated specimens was also investigated. The fracture surfaces of both unirradiated and irradiated specimens were examined. Results of these investigations are reported in the present report. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure. (au) 6 tabs., 24 ills., 9 refs

Structure and properties of UV-irradiated LLDPE and alloy of PA66 and the irradiated LLDPE

International Nuclear Information System (INIS)

Ran Qianping; Zou Hua; Wu Shishan; Shen Jian

2006-01-01

Some oxygen-containing groups such as C=O, C-O and -OH were introduced onto linear low-density polyethylene (LLDPE) chains by UV irradiation in air. Their concentration increased with the irradiation time. Crystal shape of the irradiated LLDPE remained an orthorhombic structure, while space of the crystalline plane kept unchanged. The melting temperature and crystallinity decreased due to the LLDPE chain scission into small molecules compound and crystalline defects caused by UV irradiation. Compared with pristine LLDPE, hydrophilicity of the irradiated LLDPE increased due to the introduction of polar oxygen-containing groups, but the tensile strength decreased due to the LLDPE chain degradation and reduction of crystallinity. The temperature of initial weight loss of the irradiated LLDPE was lower than that of pristine LLDPE. An alloy of PA66 and the irradiated LLDPE (irradiated PA66/LLDPE) was prepared by melting blend at 260-270 degree C. Compared to non-irradiated PA66/LLDPE alloy, dispersion of LLDPE particles in the irradiated PA66/LLDPE alloy and interfacial interactions between the components were markedly improved. Therefore, tensile strength and impact strength of the irradiated PA66/ LLDPE were higher than those of the control. (authors)

Handbook for tensile properties of austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

Database system of nuclear materials has not been developed and the physical and mechanical properties of materials used in nuclear power plant are not summarized systematically in Korea. Although Korea designs nuclear power plant, many materials used in nuclear power plant are imported because we do not have database system of nuclear material yet and it was hard to select a proper material for the structural materials of nuclear power plant. To develop database system of nuclear materials, data of mechanical, corrosion, irradiation properties are needed. Of theses properties, tensile properties are tested and summarized in this report. Tensile properties of stainless steel used in nuclear reactor internal were investigated. Data between Korea Atomic Energy Research Institute and foreign laboratory were compared to determine the precision of the result. To develope database system, materials, chemical composition, heat treatment, manufacturing process, and grain size were classified. Tensile properties were tested and summarized to use input data of database system. 9 figs., 9 tabs. (Author)

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

International Nuclear Information System (INIS)

Byun, T.S.

2001-01-01

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

An Evaluation of Fracture Toughness of Vinyl Ester Composites Cured under Microwave Conditions

Science.gov (United States)

Ku, H.; Chan, W. L.; Trada, M.; Baddeley, D.

2007-12-01

The shrinkage of vinyl ester particulate composites has been reduced by curing the resins under microwave conditions. The reduction in the shrinkage of the resins by microwaves will enable the manufacture of large vinyl ester composite items possible (H.S. Ku, G. Van Erp, J.A.R. Ball, and S. Ayers, Shrinkage Reduction of Thermoset Fibre Composites during Hardening using Microwaves Irradiation for Curing, Proceedings, Second World Engineering Congress, Kuching, Malaysia, 2002a, 22-25 July, p 177-182; H.S. Ku, Risks Involved in Curing Vinyl Ester Resins Using Microwaves Irradiation. J. Mater. Synth. Proces. 2002b, 10(2), p 97-106; S.H. Ku, Curing Vinyl Ester Particle Reinforced Composites Using Microwaves. J. Comp. Mater., (2003a), 37(22), p 2027-2042; S.H. Ku and E. Siores, Shrinkage Reduction of Thermoset Matrix Particle Reinforced Composites During Hardening Using Microwaves Irradiation, Trans. Hong Kong Inst. Eng., 2004, 11(3), p 29-34). In tensile tests, the yield strengths of samples cured under microwave conditions obtained are within 5% of those obtained by ambient curing; it is also found that with 180 W microwave power, the tensile strengths obtained for all duration of exposure to microwaves are also within the 5% of those obtained by ambient curing. While, with 360 W microwave power, the tensile strengths obtained for all duration of exposure to microwaves are 5% higher than those obtained by ambient curing. Whereas, with 540 W microwave power, the tensile strengths obtained for most samples are 5% below those obtained by ambient curing (H. Ku, V.C. Puttgunta, and M. Trada, Young’s Modulus of Vinyl Ester Composites Cured by Microwave Irradiation: Preliminary Results, J. Electromagnet. Waves Appl., 2007, 20(14), p. 1911-1924). This project, using 33% by weight fly ash reinforced vinyl ester composite [VE/FLYSH (33%)], is to further investigate the difference in fracture toughness between microwave cured vinyl ester particulate composites and those cured

Effects of electron irradiation on LDPE/MWCNT composites

Energy Technology Data Exchange (ETDEWEB)

Yang, Jianqun [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Li, Xingji, E-mail: lxj0218@hit.edu.cn [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Liu, Chaoming; Rui, Erming [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Wang, Liqin [School of Mechatronics, Harbin Institute of Technology, Harbin 150001 (China)

2015-12-15

In this study, mutiwalled carbon nanotubes (MWCNTs) were incorporated into low density polyethylene (LDPE) in different concentrations (2%, 4% and 8%) using a melt blending process. Structural, thermal stability and tensile property of the unirradiated/irradiated LDPE/MWCNT composites by 110 keV electrons were investigated by means of scanning electron microscopy (SEM), small angle X-ray scattering (SAXS), Raman spectroscopy, electron paramagnetic resonance (EPR) spectroscopy, thermogravimetric analysis (TGA) and uniaxial tensile techniques. Experimental results show that the addition of MWCNTs obviously increases the ultimate tensile strength of LDPE and decreases the elongation at break, which is attributed to the homogeneous distribution of the MWCNTs in LDPE and intense interaction between MWCNTs and LDPE matrix. Also, the electron irradiation further increases the ultimate tensile strength of LDPE/MWCNT composites, which can be ascribed to the more intense interaction between MWCNTs and LDPE matrix, and the formation of crosslinking sites in LDPE matrix induced by the electron irradiation. The addition of MWCNTs significantly enhances thermal stability of the LDPE due to the hindering effect and the scavenging free radicals, while the electron irradiation decreases thermal stability of the LDPE/MWCNT composites since the structure of the MWCNTs and LDPE matrix damages.

High-temperature irradiation effects on mechnical properties of HTGR graphites

International Nuclear Information System (INIS)

Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

1978-04-01

The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

Energy Technology Data Exchange (ETDEWEB)

Onimus, F

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Effects of reclaimed asphalt pavement on indirect tensile strength test of foamed asphalt mix tested in dry condition

International Nuclear Information System (INIS)

Katman, Herda Yati; Norhisham, Shuhairy; Ismail, Norlela; Ibrahim, Mohd Rasdan; Matori, Mohd Yazip

2013-01-01

Indirect tensile strength (ITS) test was conducted to analyse strength of the foamed asphalt mixes incorporating reclaimed asphalt pavement. Samples were tested for ITS after cured in the oven at 40°C for 72 hours. This testing condition known as dry condition or unconditioned. Laboratory results show that reclaimed asphalt pavement (RAP) contents insignificantly affect the ITS results. ITS results significantly affected by foamed bitumen contents.

Influence of irradiation conditions on the gamma irradiation effect in polyethylene

International Nuclear Information System (INIS)

Kacarevic-Popovic, Z.; Gal, O.; Novakovic, L.J.; Secerov, B.

2002-01-01

Complete text of publication follows. The radiation cross-linking of polyethylene, due to its high cross-linking yield, has resulted in the radiation technology that has found application in radiation production of heat shrinkable structures and in improvement of mechanical and thermo-physical properties of oriented polyethylene objects. It is observed that the cross-linking efficiency decreases when the irradiation is carried out in the presence of oxygen. In order to estimate the conditions that improve cross-linking efficiency, gamma irradiation effect in two types of polyethylene, irradiated in water and air was investigated. The polyethylene samples used were the low density (LDPE) Lotrene CdF 0302 with 45% crystallinity and the high density (HDPE) Hiplex EHM 6003 with 73% crystallinity. Both kinds of samples, fixed in the Pyrex glass tubes, were simultaneously irradiated with 60 Co gamma rays in distilled water and air, at a doses rate of 9,5 kGy/h (determined by the Fricke dosimeter) at room temperature. Radiation induced oxidative degradation was followed through oxygen containing group formation by the carbonyl group band (1720 cm -1 ) and transvinylene group formation by the band at 966 cm -1 in the infrared spectra. Cross-linking efficiency was determined by gel content using the procedure of the extraction in xylene. The monitored effects of gamma irradiation in water and air point to the conclusion that irradiation in water leads to the lower oxidative degradation and higher cross-linking compared with the effects measured after irradiation in air

Material property changes of stainless steels under PWR irradiation

International Nuclear Information System (INIS)

Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

2009-01-01

Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

Deformation twinning in irradiated ferritic/martensitic steels

Science.gov (United States)

Wang, K.; Dai, Y.; Spätig, P.

2018-04-01

Two different ferritic/martensitic steels were tensile tested to gain insight into the mechanisms of embrittlement induced by the combined effects of displacement damage and helium after proton/neutron irradiation in SINQ, the Swiss spallation neutron source. The irradiation conditions were in the range: 15.8-19.8 dpa (displacement per atom) with 1370-1750 appm He at 245-300 °C. All the samples fractured in brittle mode with intergranular or cleavage fracture surfaces when tested at room temperature (RT) or 300 °C. After tensile test, transmission electron microscopy (TEM) was employed to investigate the deformation microstructures. TEM-lamella samples were extracted directly below the intergranular fracture surfaces or cleavage surfaces by using the focused ion beam technique. Deformation twinning was observed in irradiated specimens at high irradiation dose. Only twins with {112} plane were observed in all of the samples. The average thickness of twins is about 40 nm. Twins initiated at the fracture surface, became gradually thinner with distance away from the fracture surface and finally stopped in the matrix. Novel features such as twin-precipitate interactions, twin-grain boundary and/or twin-lath boundary interactions were observed. Twinning bands were seen to be arrested by grain boundaries or large precipitates, but could penetrate martensitic lath boundaries. Unlike the case of defect free channels, small defect-clusters, dislocation loops and dense small helium bubbles were observed inside twins.

The development of a tensile-shear punch correlation for yield properties of model austenitic alloys

Energy Technology Data Exchange (ETDEWEB)

Hankin, G.L.; Faulkner, R.G. [Loughborough Univ. (United Kingdom); Hamilton, M.L.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

1997-08-01

The effective shear yield and maximum strengths of a set of neutron-irradiated, isotopically tailored austentic alloys were evaluated using the shear punch test. The dependence on composition and neutron dose showed the same trends as were observed in the corresponding miniature tensile specimen study conducted earlier. A single tensile-shear punch correlation was developed for the three alloys in which the maximum shear stress or Tresca criterion was successfully applied to predict the slope. The correlation will predict the tensile yield strength of the three different austenitic alloys tested to within {+-}53 MPa. The accuracy of the correlation improves with increasing material strength, to within {+-} MPa for predicting tensile yield strengths in the range of 400-800 MPa.

Effects of non-steady irradiation conditions on fusion materials performance

International Nuclear Information System (INIS)

Matsui, H.; Fukumoto, K.; Nagumo, T.; Nita, N.

2001-01-01

During startup of fusion reactors, materials are exposed to neutron irradiation under non-steady temperature condition. Since the temperature of irradiation has decisive effects on the microstructural evolution, the non-steady temperature will have important consequences in the performance of fusion reactor materials. In the present study, a series of vanadium based alloys have been irradiated with neutrons in a temperature cycling condition. It has been found from this study that cavity number density is much greater in temperature cycled specimens than in steady temperature irradiation. Keeping the upper temperature constant, cavity number density is greater for smaller difference between the upper and the lower temperature. It follows that relatively small temperature excursions may have rather significant effects on the fusion material performance in service. (author)

Behaviour of the activity of cellulase irradiated under various conditions

International Nuclear Information System (INIS)

Kumakura, M.; Kaetsu, I.

1988-01-01

The activity of cellulase irradiated at various conditions has been studied. The activity of cellulase irradiated at low temperature (-78 0 C) increased by heating at 40 0 C, but that of cellulase irradiated at high temperature above 0 0 C decreased. The activity of cellulase irradiated in the dry state at room temperature increased with irradiation dose. The effect of adding biological substances such as amino acids, enzymes, and agar on the irradiation of cellulase was studied. It was shown that EDTA and p-benzoquinone have a protective ability against radiation-induced inactivation of the enzyme. (author)

Behaviour of the activity of cellulase irradiated under various conditions

Energy Technology Data Exchange (ETDEWEB)

Kumakura, M; Kaetsu, I

1988-04-18

The activity of cellulase irradiated at various conditions has been studied. The activity of cellulase irradiated at low temperature (-78 /sup 0/C) increased by heating at 40 /sup 0/C, but that of cellulase irradiated at high temperature above 0 /sup 0/C decreased. The activity of cellulase irradiated in the dry state at room temperature increased with irradiation dose. The effect of adding biological substances such as amino acids, enzymes, and agar on the irradiation of cellulase was studied. It was shown that EDTA and p-benzoquinone have a protective ability against radiation-induced inactivation of the enzyme.

Tensile testing

CERN Document Server

2004-01-01

A complete guide to the uniaxial tensile test, the cornerstone test for determining the mechanical properties of materials: Learn ways to predict material behavior through tensile testing. Learn how to test metals, alloys, composites, ceramics, and plastics to determine strength, ductility and elastic/plastic deformation. A must for laboratory managers, technicians, materials and design engineers, and students involved with uniaxial tensile testing. Tensile Testing , Second Edition begins with an introduction and overview of the test, with clear explanations of how materials properties are determined from test results. Subsequent sections illustrate how knowledge gained through tensile tests, such as tension properties to predict the behavior (including strength, ductility, elastic or plastic deformation, tensile and yield strengths) have resulted in improvements in materals applications. The Second Edition is completely revised and updated. It includes expanded coverage throughout the volume on a variety of ...

The effect of neutron irradiation on the mechanical properties of welded zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Evans, D G

1962-07-15

Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260{sup o}C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900{sup o}C for 40 minutes, another group contained welded areas in the centre of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld, Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. The transition temperature of Zircaloy-2 increased appreciably upon welding. This was accompanied by a decrease in absorbed energy values for all temperatures between 0{sup o} and 300{sup o}C. Neutron irradiation had no effect on the impact properties of welded. Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260{sup o}C, and increased the 260{sup o}C PL, YS and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260{sup o}C strength properties of the as-welded material. It was found that the unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. These fractures were similar to those observed in irradiated fuel bundles which had been damaged during transfer operations. A large amount of scatter rendered some

Effects of low doses of 14-MeV neutrons on the tensile properties of three binary copper alloys

International Nuclear Information System (INIS)

Heinisch, H.L.; Pintler, J.S.

1986-12-01

Miniature tensile specimens of high purity copper and copper alloyed respectively with five atom percent of Al, Mn, and Ni were irradiated with D-T fusion neutrons in the RTNS-II to fluences up to 1.3 x 10 18 n/cm 2 at 90 0 C. To compare fission and fusion neutron effects, some specimens were also irradiated at the same temperature to similar damage levels in the Omega West Reactor (OWR). Tensile tests were performed at room temperature, and the radiation-induced changes in tensile properties were examined as functions of displacements per atom (dpa). The irradiation-induced strengthening of Cu5%Mn is greater than that of Cu5%Al and Cu5%Ni, which behave about the same. However, all the alloys sustain less irradiation-induced strengthening by 14 MeV neutrons than pure copper, which is in contrast to the reported results of earlier work using hardness measurements. The effects of fission and fusion neutrons on the yield stress of Cu5%Al and Cu5%Ni correlate well on the basis of dpa, but the data for Cu5%Mn suggest that dpa may not be a good correlation parameter for this alloy in this fluence and temperature range

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

The development of a tensile-shear punch correlation for yield properties of model austenitic alloys

International Nuclear Information System (INIS)

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

1997-01-01

The effective shear yield and maximum strengths of a set of neutron-irradiated, isotopically tailored austentic alloys were evaluated using the shear punch test. The dependence on composition and neutron dose showed the same trends as were observed in the corresponding miniature tensile specimen study conducted earlier. A single tensile-shear punch correlation was developed for the three alloys in which the maximum shear stress or Tresca criterion was successfully applied to predict the slope. The correlation will predict the tensile yield strength of the three different austenitic alloys tested to within ±53 MPa. The accuracy of the correlation improves with increasing material strength, to within ± MPa for predicting tensile yield strengths in the range of 400-800 MPa

Revised ANL-reported tensile data for V-Ti and V-Cr-Ti alloys

Energy Technology Data Exchange (ETDEWEB)

Billone, M.C. [Argonne National Lab., IL (United States)

1997-08-01

The tensile for all irradiated vanadium alloy samples and several unirradiated vanadium alloys tested at Argonne National Laboratory (ANL) have been critically reviewed and revised, as necessary. The review and revision are based on re-analyzing the original load-displacement strip-chart recording using a methodology consistent with current ASTM standards. No significant difference has been found between the newly-revised and previously-reported values of yield strength (YS) and ultimate tensile strength (UTS). However, by correctly subtracting the non-gauge-length displacement and linear gauge-length displacement from the total cross-head displacement, the uniform elongation (UE) of the gauge length decreases by 4-9% strain and the total elongation (TE) of the gauge length decreases by 1-7% strain. These differences are more significant for lower-ductility irradiated alloys than for higher-ductility alloys.

Revised ANL-reported tensile data for V-Ti and V-Cr-Ti alloys

International Nuclear Information System (INIS)

Billone, M.C.

1997-01-01

The tensile for all irradiated vanadium alloy samples and several unirradiated vanadium alloys tested at Argonne National Laboratory (ANL) have been critically reviewed and revised, as necessary. The review and revision are based on re-analyzing the original load-displacement strip-chart recording using a methodology consistent with current ASTM standards. No significant difference has been found between the newly-revised and previously-reported values of yield strength (YS) and ultimate tensile strength (UTS). However, by correctly subtracting the non-gauge-length displacement and linear gauge-length displacement from the total cross-head displacement, the uniform elongation (UE) of the gauge length decreases by 4-9% strain and the total elongation (TE) of the gauge length decreases by 1-7% strain. These differences are more significant for lower-ductility irradiated alloys than for higher-ductility alloys

Irradiation effects in tungsten-copper laminate composite

Energy Technology Data Exchange (ETDEWEB)

Garrison, L.M., E-mail: garrisonlm@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Byun, T.S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Reiser, J.; Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany)

2016-12-01

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410–780 °C and fast neutron fluences of 0.02–9.0 × 10{sup 25} n/m{sup 2}, E > 0.1 MeV, 0.0039–1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22 °C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22 °C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile. - Highlights: • Fusion reactors need a tough, ductile tungsten plasma-facing material. • The unirradiated tungsten-copper laminate is more ductile than tungsten alone. • After neutron irradiation, the composite has significantly less ductility. • The tungsten behavior appears to dominate the overall composite behavior.

Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

International Nuclear Information System (INIS)

Watanabe, Katsutoshi

1987-09-01

The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

Measurements of three dimensional residual stress distribution on laser irradiated spot

International Nuclear Information System (INIS)

Tanaka, Hirotomo; Akita, Koichi; Ohya, Shin-ichi; Sano, Yuji; Naito, Hideki

2004-01-01

Three dimensional residual stress distributions on laser irradiated spots were measured using synchrotron radiation to study the basic mechanism of laser peening. A water-immersed sample of high tensile strength steel was irradiated with Q-switched and frequency-doubled Nd:YAG laser. The residual stress depth profile of the sample was obtained by alternately repeating the measurement and surface layer removal by electrolytic polishing. Tensile residual stresses were observed on the surface of all irradiated spots, whereas residual stress changed to compressive just beneath the surface. The depth of compressive residual stress imparted by laser irradiation and plastic deformation zone increased with increasing the number of laser pulses irradiated on the same spot. (author)

Evaluation of the tensile bond strength of an adhesive system self-etching in dentin irradiated with Er:YAG laser; Avaliacao da resistencia a tracao de um sistema adesivo self-etching em dentina irradiada com Er:YAG laser

Energy Technology Data Exchange (ETDEWEB)

Mello, Andrea Malluf Dabul de

2000-07-01

Since Buonocore (1955), several researchers have been seeking for the best adhesive system and treatment for the enamel and dentin surfaces. The use of the acid has been presented as one of the best techniques of dentin conditioning , because this promotes the removal of the 'smear layer and exhibition of dentinal structure, for a best penetration and micro- retention of the adhesive system. However, some conditioning methods have been appearing in the literature, for the substitution or interaction with the acid substances, as the laser. The objective of this work is to evaluate the tensile bond strength of the adhesive system self-etching' associated to a composed resin, in dentin surfaces conditioned with the Er:YAG laser. For this study, freshly extracted human teeth were used and in each one the dentinal surfaces , which were treated with three sandpapers of different granulations (120,400,600), to obtain a standard of the smear layer, before the irradiation of the laser and of the restoring procedure. After these procedures the specimens were storage in distilled water at 37 deg C for 24 hours. Soon after, they were submitted to the tensile strength test .After analyzing the results, we can concluded that the use of the Er:YAG laser can substitute the drill without the need of conditioning, when using the adhesive system 'self-etching' in the dentinal surfaces because there was a decline in the strength of adhesion in the groups conditioned with the laser. (author)

Effect of Spark-Plasma-Sintering Conditions on Tensile Properties of Aluminum Matrix Composites Reinforced with Multiwalled Carbon Nanotubes (MWCNTs)

Science.gov (United States)

Chen, B.; Imai, H.; Umeda, J.; Takahashi, M.; Kondoh, K.

2017-04-01

In this study, aluminum (Al) matrix composites containing 2 wt.% multiwalled carbon nanotubes (CNTs) were fabricated by powder metallurgy using high-energy ball milling (HEBM), spark plasma sintering (SPS), and subsequent hot extrusion. The effect of SPS conditions on the tensile properties of CNT/Al composites was investigated. The results showed that composites with well-dispersed CNTs and nearly full-density CNT/Al can be obtained. During HEBM, CNTs were shortened, inserted into welded Al powder particles, bonded to Al, and still stable without CNT-Al reaction. After consolidation, Al4C3 phases formed in composites under different sintering conditions. With the increase of sintering temperature and holding time, the strength decreased. Conversely, the ductility and toughness noticeably increased. As a result, a good balance between strength (367 MPa in ultimate tensile strength) and ductility (13% in elongation) was achieved in the as-extruded CNT/Al composite sintered at 630°C with a holding time of 300 min.

Effects of CTR irradiation on the mechanical properties of structural materials

International Nuclear Information System (INIS)

Wiffen, F.W.

1976-11-01

Mechanical properties of CTR structural materials are important in determining the reliability and economics of fusion power. Furthermore, these properties are significantly affected by the high neutron flux experienced by components in the regions near the plasma of the fusion reactor. In general, irradiation hardens the material and leads to a reduction in ductility. An exception to this is in some complex engineering alloys where either hardening or softening can be observed depending on the alloy and the irradiation conditions. Regardless of this restriction, irradiation usually leads to a reduction in ductility. Available tensile data examined in this paper show that significant ductility reduction can be found for irradiation conditions typical of CTR operation. Consideration of these effects show that extensive work will be needed to fully establish the in-service properties of CTR structures. This information will be used by designers to develop conditions and design philosophies adapted to avoid the most deleterious conditions and minimize stresses on structures on reactor design. The information will also be used as input to alloy development programs with goals of producing materials more resistant to property degradation during irradiation. It is clear that a great deal of additional work will be required both to understand the effect of CTR irradiation on properties and to develop optimal alloys for this application

Development of the transverse tensile and fracture toughness test techniques for spent fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Ahn, S. B.; Hong, K. P.; Jung, Y. H.; Seo, H. S.; Oh, W. H.; Yoo, B. O.; Kim, D. S.; Seo, K. S

2001-12-01

To define the cause of cladding damage which can take place during the operation of nuclear power plant and the storage through the degradation aspect of mechanical characteristics, the transverse tensile an fracture toughness test were developed in hot cell at IMEF(Irradiated Material Experiment Facility). The following hot cell techniques were developed. 1. The development of a jig and a specimen for transverse tensile test 2. The acquisition of a manufacturing technique for the transverse tensile specimen at hot cell 3. The acquisition of testing procedures and an analysis technque for the transverse tensile 4. The dimensional determination of an optimized fracture toughness specimen 5. The acquisition of manufacturing technique for the fracture toughness test specimen at the hot cell 6. The acquisition of testing procedures and analysis technique for the fracture toughness test (Multiple specimen method, DCPD method, Load ratio method)

Improvement of life time of SCC in type 304 stainless steel by ultrasound irradiation

International Nuclear Information System (INIS)

Tokiwai, Moriyasu; Kimura, Hideo

1985-01-01

It is well known that the susceptibility to stress corrosion cracking (SCC) is controled by compressive stress such as shot-peening treatment. In this study, the effects of ultrasound irradiation to type 304 stainless upon SCC were investigated. The main findings are as follows; (1) Ultrasound irradiation produces the high level compressive stress on the surface of metals. This compressive stress was induced by the cavitation phenomenon. (2) In U-bent specimen, the initial tensile stress was mitigated and converted to compressive stress by ultrasound irradiation. (3) Type 304 stainless steel was subjected to SCC test using sodium thyosulfate solution. It was definitely demonstrated that the ultrasound irradiation was effective for the mitigation of SCC life time. (4) Ultrasound irradiation time was one of the most important factors in irradiation conditions. (author)

Correlation between electron-irradiation defects and applied stress in graphene: A molecular dynamics study

Energy Technology Data Exchange (ETDEWEB)

Kida, Shogo; Yamamoto, Masaya; Kawata, Hiroaki; Hirai, Yoshihiko; Yasuda, Masaaki, E-mail: yasuda@pe.osakafu-u.ac.jp [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Tada, Kazuhiro [Department of Electrical and Control Systems Engineering, National Institute of Technology, Toyama College, Toyama 939-8630 (Japan)

2015-09-15

Molecular dynamics (MD) simulations are performed to study the correlation between electron irradiation defects and applied stress in graphene. The electron irradiation effect is introduced by the binary collision model in the MD simulation. By applying a tensile stress to graphene, the number of adatom-vacancy (AV) and Stone–Wales (SW) defects increase under electron irradiation, while the number of single-vacancy defects is not noticeably affected by the applied stress. Both the activation and formation energies of an AV defect and the activation energy of an SW defect decrease when a tensile stress is applied to graphene. Applying tensile stress also relaxes the compression stress associated with SW defect formation. These effects induced by the applied stress cause the increase in AV and SW defect formation under electron irradiation.

Mechanical properties of 1950's vintage 304 stainless steel weldment components after low temperature neutron irradiation

International Nuclear Information System (INIS)

Sindelar, R.L.; Caskey, G.R. Jr.; Thomas, J.K.; Hawthorne, J.R.; Hiser, A.L.; Lott, R.A.; Begley, J.A.; Shogan, R.P.

1991-01-01

The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950's from Type 304 stainless steel plates welded with Type 308 stainless steel filler using the multipass metal inert gas process. An irradiated mechanical properties database has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150 degrees C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25 degrees C and 125 degrees C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125 degrees C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials. The HFIR-irradiated materials show an increase in yield strength between about 20% and 190% with a concomitant tensile strength increase between about 15% to 30%. The elastic-plastic fracture toughness parameters and Charpy-V energy absorption both decrease and show only a slight sensitivity to dose. The irradiation-induced decrease in the elastic-plastic fracture toughness (J def at 1 mm crack extension) is between 20% to 65%; the range of J 1C values are 72.8 to 366 kJ/m 2 for the irradiated materials. Similarly, Charpy V-notch results show a 40% to 60% decrease in impact energies

The Effect of Gamma Irradiation on Physico-Chemical Properties of Sodium Alginate

International Nuclear Information System (INIS)

Erizal; Sudrajat, A.; Dewi SP; Rahayu Chosdu; Tatiek Martati

2004-01-01

In the order to develop application of radiation to improve the quality of natural polymers, irradiation effects on physico-chemical characterization of Na-alginate has been carried out. Both Na-alginate powder and solution were irradiated by gamma rays at doses of 0;10;30; and 50 kGy. The parameters observed consisting of viscosity, pH, tensile strength, IR and UV spectra of irradiated and unirradiated samples. The results showed that irradiation up to 50 kGy, did not give effect on pH and no new peak appeared in the IR spectra. However, tensile strength and viscosity decreased with increasing irradiation dose up 50 kGy, which showed the possibility of degradation process on Na-alginate. (author)

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

International Nuclear Information System (INIS)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430 0 C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by a factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material

PDS 1-5. Divertor heat sink materials pre- and post-neutron irradiation. Tensile and fatigue tests of brazed joints of molybdenum alloys and 316L stainless steel

International Nuclear Information System (INIS)

Lind, Anders.

1994-01-01

Tensile specimens from brazed joints of molybdenum alloys (TZM or Mo-5%Re) and Type 316L austenitic stainless steel tubes have been tested at ambient temperature and 127 degrees C before and after neutron irradiation at about 40 degrees C to approximately 0.2 dpa. The unirradiated specimens showed generally ductile behaviour, but the irradiated specimens were notch sensitive and failed in a brittle manner with zero elongation; in all cases the fracture occurred in the molybdenum alloy. The brittle behaviour is consistent with previously published data and results from the increase in strength (radiation hardening) and the associated increase in the ductile-brittle transition temperature (radiation embrittlement) induced in the body-centered-cubic (BCC) molybdenum alloys by irradiation to relatively low displacement doses. The same type of irradiated specimens were also used in fatigue tests. However, the results from the fatigue tests are too limited and complementary studies are needed. During exposure to water locally up to 25% of the wall thickness of the Mo-alloys has corroded away. These observations cast serious doubts on the viability of the molybdenum alloys for divertor applications in fusion systems. 8 refs, 29 figs

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Onimus, F.

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Effect of gamma irradiation on the sensitivities of escherichia coli at deep frozen conditions

International Nuclear Information System (INIS)

Takigami, Machiko; Ito, Hitoshi

1996-01-01

Phosphate buffer suspensions of three strains of Escherichia coli were irradiated with gamma-rays at room temperature and deep frozen conditions. They were inoculated on MacConkey agar plates to see the comparative sensitivities to the irradiation. Compared to the irradiation at room temperature, the sensitivities of the strains decreased by irradiation at deep frozen conditions. Addition of glycerol to the E. coli suspensions decreased the sensitivities of E. coli to gamma-rays not only at room temperature but also at deep frozen conditions. These phenomena were elucidated by the decrease of production and mobility of OH radicals at deep frozen conditions. (author)

New Scanning Electron Microscope Used for Cryogenic Tensile Testing

CERN Multimedia

Maximilien Brice

2013-01-01

At CERN engineering department's installation for cryogenic tensile testing, the new scanning electron microscope (SEM) allows for detailed optical observations to be carried out. Using the SEM, surface coatings and tensile properties of materials can investigated in order to better understand how they behave under different conditions.

Post Irradiation Mechanical Behaviour of Three EUROFER Joints

International Nuclear Information System (INIS)

Lucon, E.; Leenaers, A.; Vandermeulen, W.

2006-01-01

The post-irradiation mechanical properties of three EUROFER joints (two diffusion joints and one TIG weld) have been characterized after irradiation to 1.8 dpa at 300 degrees Celsius in the BR-2 reactor. Tensile, KLST impact and fracture toughness tests have been performed. Based on the results obtained and on the comparison with data from EUROFER base material irradiated under similar conditions, the post-irradiation mechanical behaviour of both diffusion joints (laboratory and mock-up) appears similar to that of the base material. The properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region. Thus, specimens from the upper layer exhibit extremely pronounced hardening and embrittlement caused by irradiation. The samples extracted from the lower layer show much better resistance to neutron exposure, although their measured properties do not match those of the diffusion joints. The results presented demonstrate that diffusion joining can be a very promising technique.

Irradiation effects on aluminium and beryllium

International Nuclear Information System (INIS)

Bieth, M.

1992-01-01

The High Flux Reactor (HFR) in Petten (The Netherlands) is a 45 MW light water cooled and moderated research reactor. The vessel was replaced in 1984 after more than 20 years of operation because doubts had arisen over the condition of the aluminium alloy construction material. Data on the mechanical properties of the aluminium alloy Al 5154 with and without neutron irradiation are necessary for the safety analysis of the new HFR vessel which is constructed from the same material as the old vessel. Fatigue, fracture mechanics (crack growth and fracture toughness) and tensile properties have been obtained from several experimental testing programmes with materials of the new and the old HFR vessel. 1) Low-cycle fatigue testing has been carried out on non-irradiated specimens from stock material of the new HFR vessel. The number of cycles to failure ranges from 90 to more than 50,000 for applied strain from 3.0% to 0.4%; 2) Fatigue crack growth rate testing has been conducted: - with unirradiated specimens from stock material of the new vessel; - with irradiated specimens from the remnants of the old core box. Irradiation has a minor effect on the sub-critical fatigue crack growth rate. The ultimate increase of the mean crack growth rate amounts to a factor of 2. However crack extension is strongly reduced due to the smaller crack length for crack growth instability (reduction of K IC ). - Irradiated material from the core box walls of the old vessel has been used for fracture toughness testing. The conditional fracture toughness values K IQ ranges from 30.3 down to 16.5 MPa√m. The lowermost meaningful 'K IC ' is 17.7 MPa√m corresponding to the thermal fluence of 7.5 10 26 n/m 2 for the End of Life (EOL) of the old vessel. - Testing carried out on irradiated material from the remnants of the old HFR core box shows an ultimate neutron irradiation hardening of 35 points increase of HSR 15N and an ultimate tensile yield stress of 589 MPa corresponding to the

Supercritical CO2 foaming of radiation crosslinked polypropylene/high-density polyethylene blend: Cell structure and tensile property

Science.gov (United States)

Yang, Chenguang; Xing, Zhe; Zhang, Mingxing; Zhao, Quan; Wang, Mouhua; Wu, Guozhong

2017-12-01

A blend of isotactic polypropylene (PP) with high-density polyethylene (HDPE) in different PP/HDPE ratios was irradiated by γ-ray to induce cross-linking and then foamed using supercritical carbon dioxide (scCO2) as a blowing agent. Radiation effect on the melting point and crystallinity were analyzed in detail. The average cell diameter and cell density were compared for PP/HDPE foams prepared under different conditions. The optimum absorbed dose for the scCO2 foaming of PP/HDPE in terms of foaming ability and cell structure was 20 kGy. Tensile measurements showed that the elongation at break and tensile strength at break of the crosslinked PP/HDPE foams were higher than the non-crosslinked ones. Of particular interest was the increase in the foaming temperature window from 4 ℃ for pristine PP to 8-12 ℃ for the radiation crosslinked PP/HDPE blends. This implies much easier handling of scCO2 foaming of crosslinked PP with the addition of HDPE.

Effects of irradiation at low temperature on V-4Cr-4Ti

International Nuclear Information System (INIS)

Alexander, D.J.; Snead, L.L.; Zinkle, S.J.

1996-01-01

Irradiation at low temperatures (100 to 275 degrees C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275 degrees C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation

Effects of irradiation at low temperature on V-4Cr-4Ti

Energy Technology Data Exchange (ETDEWEB)

Alexander, D.J.; Snead, L.L.; Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

1996-10-01

Irradiation at low temperatures (100 to 275{degrees}C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275{degrees}C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation.

Inert medium (helium) irradiation testing of pressure tube samples

International Nuclear Information System (INIS)

Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

2001-01-01

Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

Spectroscopic analysis of radiation-generated changes in tensile properties of a polyetherimide film

International Nuclear Information System (INIS)

Long, E.R. Jr.; Long, S.A.T.

1985-05-01

The effects of electron radiation on Ultem, a polyetherimide were studied for doses from 2 x 10 to the 9th power to 6 x 10 to the 9th power rad. Specimens were studied for tensile property testing and for electron paramagnetic resonance and infrared spectroscopic measurements of molecular structure. A Faraday cup design and a method for remote temperature measurement were developed. The spectroscopic data show that radiation caused dehydrogenation of methyl groups, rupture of main-chain ether linkage, and opening of imide rings, all to form radicals and indicate that the so-formed atomic hydrogen attached to phenyl radicals, but not to phenoxyl radicals, which would have formed hydroxyls. The observed decays of the radiation-generated phenoxyl, gem-dimethyl, and carbonyl radicals were interpreted as a combining of the radicals to form crosslinking. This crosslinking is the probable cause of the major reduction in the elongation of the tensile specimens after irradiation. Subsequent classical solubility tests indicate that the irradiation caused massive crosslinking

Microstructures of beta-silicon carbide after irradiation creep deformation at elevated temperatures

International Nuclear Information System (INIS)

Katoh, Yutai; Kondo, Sosuke; Snead, Lance L.

2008-01-01

Microstructures of silicon carbide were examined by transmission electron microscopy (TEM) after creep deformation under neutron irradiation. Thin strip specimens of polycrystalline and monocrystalline, chemically vapor-deposited, beta-phase silicon carbide were irradiated in the high flux isotope reactor to 0.7-4.2 dpa at nominal temperatures of 640-1080 deg. C in an elastically pre-strained bend stress relaxation configuration with the initial stress of ∼100 MPa. Irradiation creep caused permanent strains of 0.6 to 2.3 x 10 -4 . Tensile-loaded near-surface portions of the crept specimens were examined by TEM. The main microstructural features observed were dislocation loops in all samples, and appeared similar to those observed in samples irradiated in non-stressed conditions. Slight but statistically significant anisotropy in dislocation loop microstructure was observed in one irradiation condition, and accounted for at least a fraction of the creep strain derived from the stress relaxation. The estimated total volume of loops accounted for 10-45% of the estimated total swelling. The results imply that the early irradiation creep deformation of SiC observed in this work was driven by anisotropic evolutions of extrinsic dislocation loops and matrix defects with undetectable sizes

Tensile behavior and tension stiffening of reinforced concrete

International Nuclear Information System (INIS)

Choun, Young Sun; Seo, Jeong Moon

2001-03-01

For the ultimate behavior analysis of containment buildings under severe accident conditions, a clear understanding of tensile behaviors of plain and reinforced concrete is necessary. Nonlinear models for tensile behaviors of concrete are also needed. This report describe following items: tensile behaviors of plain concrete, test results of reinforced concrete panels in uniaxial and biaxial tension, tension stiffening. The tensile behaviors of reinforced concrete are significantly influenced by the properties of concrete and reinforcing steel. Thus, for a more reliable evaluation of tensile behavior and ultimate pressure capacity of a reinforced or prestressed concrete containment building, an advanced concrete model which can be considered rebar-concrete interaction effects should be developed. In additions, a crack behavior analysis method and tension stiffening models, which are based on fracture mechanics, should be developed. The model should be based on the various test data from specimens considering material and sectional properties of the containment building

Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

International Nuclear Information System (INIS)

Vladimirov, P.; Moeslang, A.

2004-01-01

Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

Improvement of carbon fibre surface properties using electron beam irradiation

International Nuclear Information System (INIS)

Eddy Segura Pino; Luci Diva Brocardo Machado; Claudia Giovedi

2006-01-01

Carbon fiber-reinforced advance composites have been used for structural applications, mainly due to their mechanical properties, and additional features such as high strength-to-weight ratio, stiffness-to-weight ratio, corrosion resistance and wear properties. The main factor for a good mechanical performance of carbon fiber-reinforced composite is the interfacial interaction between the components that are fiber and polymeric matrix. The greatest challenge is to improve adhesion between components having elasticity modulus which differ by orders of magnitude and furthermore they are immiscible in each other. Another important factor is the sizing material on the carbon fiber, which protects the carbon fiber filaments and must be compatible with the matrix material in order to improve the adhesion process. The interaction of ionizing radiation from electron beam can induce in the irradiated material the formation of very active centers and free radicals. Further evolution of these active species can significantly modify structure and properties not only in the irradiated polymeric matrix but also on the fiber surface. So that, fiber and matrix play an important role in the production of chemical bonds, which promote better adhesion between both materials improving the composite mechanical performance. The aim of this work was to improve the surface properties of the carbon fiber surface using ionizing radiation from an electron beam in order to obtain improvement of the adhesion properties in the resulted composite. Commercial carbon fiber roving of high tensile strength with 12 000 filaments named 12 k, and sizing material of epoxy resin modified by ester groups was studied. EB irradiation has been carried out at the Institute for Nuclear and Energy Research (IPEN) facilities using a 1.5 MeV 37.5 kW Dynamitron electron accelerator model JOB-188. Rovings of carbon fibers with 1.78 g cm -3 density and 0.13 mm thickness were irradiated with 0.555 MeV, 6.43 mA and

Study on radiation effect of poly (vinyl alcohol) films irradiated by tritium decay

International Nuclear Information System (INIS)

Li Hairong; Peng Shuming; Zhou Xiaosong; Yu Mingming; Xia Lidong; Chen Xiaohua; Liang Jianhua

2014-01-01

The radiation effect of poly(vinyl alcohol) films used as a kind of gas-barrier material for inertial confinement fusion (ICF) targets was studied under the different conditions of β-ray from tritium decay. The changes of physical and chemical properties of the irradiated material samples were analyzed by FTIR, XRD and AFM. The tritium-hydrogen isotopic exchange reaction of the irradiated samples mainly occurs at C-H bond and the IR absorption peak of C-T bond obviously increases with the irradiation dose. For strong hydrogen bonding interaction, the isotopic exchange reaction doesn't occur at O-H bond. The crystallinity degree and surface morphology of the irradiated samples were changed. The tensile properties of irradiated poly(vinyl alcohol) films were measured by universal material testing machine. The results show that the change trend of mechanical properties is in accordance with the microstructures of the irradiated samples. (authors)

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

International Nuclear Information System (INIS)

Gelles, D.S.

1984-05-01

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

Improvement of carbon fiber surface properties using electron beam irradiation

International Nuclear Information System (INIS)

Pino, E.S.; Machado, L.D.B.; Giovedi, C.

2007-01-01

Carbon fiber-reinforced advance composites have been used for structural applications, mainly on account of their mechanical properties. The main factor for a good mechanical performance of carbon fiber-reinforced composite is the interfacial interaction between its components, which are carbon fiber and polymeric matrix. The aim of this study is to improve the surface properties of the carbon fiber using ionizing radiation from an electron beam to obtain better adhesion properties in the resultant composite. EB radiation was applied on the carbon fiber itself before preparing test specimens for the mechanical tests. Experimental results showed that EB irradiation improved the tensile strength of carbon fiber samples. The maximum value in tensile strength was reached using doses of about 250 kGy. After breakage, the morphology aspect of the tensile specimens prepared with irradiated and non-irradiated car- bon fibers were evaluated. SEM micrographs showed modifications on the carbon fiber surface. (authors)

Dose Profiles in ECAL Crystals for Various Irradiation Conditions

CERN Document Server

Huhtinen, Mika

1998-01-01

Simulated dose profiles in various irradiation and beam test conditions are compared to the expected dose profiles in the ECAL crystals at LHC. Simple front or side irradiations with photons give too steep or too flat dose profiles, respectively. Thus, if dose maxima are fitted to agree, front irradiation underestimate the average dose whereas side irradiations tend to overestimate. Different profiles are difficult to compare reliably, but it seems likely that in both cases the discrepancy is about a factor of 2-3 but in different directions. For most purposes this is likely to be good enough, but should be taken into account in the interpretation of the test results. It is shown that using a customized lead mask between the source and the crystal can significantly improve the agreement between 60 Co side irradiations and the LHC predictions. A 400 MeV/c pion beam incident on a crystal matrix can also reproduce rather well the profiles expected in the barrel ECAL.

Tensile properties of unirradiated path A PCA

International Nuclear Information System (INIS)

Braski, D.N.; Maziasz, P.J.

1983-01-01

The tensile properties of PCA in the Al (solution annealed), A3 (25%-cold worked), and B2 (aged, cold worked, and reaged) conditions were determined from room temperature to 600 0 C. The tensile behavior of PCA-A1 and -A3 was generally similar to that of titanium-modified type 316 stainless steel with similar microstructures. The PCA-B2 was weaker than PCA-A3, especially above 500 0 C, but demonstrated slightly better ducility

Radiometric ratio characterization for low-to-mid CPV modules operating in variable irradiance conditions

Science.gov (United States)

Vorndran, Shelby; Russo, Juan; Zhang, Deming; Gordon, Michael; Kostuk, Raymond

2012-10-01

In this work, a concentrating photovoltaic (CPV) design methodology is proposed which aims to maximize system efficiency for a given irradiance condition. In this technique, the acceptance angle of the system is radiometrically matched to the angular spread of the site's average irradiance conditions using a simple geometric ratio. The optical efficiency of CPV systems from flat-plate to high-concentration is plotted at all irradiance conditions. Concentrator systems are measured outdoors in various irradiance conditions to test the methodology. This modeling technique is valuable at the design stage to determine the ideal level of concentration for a CPV module. It requires only two inputs: the acceptance angle profile of the system and the site's average direct and diffuse irradiance fractions. Acceptance angle can be determined by raytracing or testing a fabricated prototype in the lab with a solar simulator. The average irradiance conditions can be found in the Typical Metrological Year (TMY3) database. Additionally, the information gained from this technique can be used to determine tracking tolerance, quantify power loss during an isolated weather event, and do more sophisticated analysis such as I-V curve simulation.

Microstructural defects in EUROFER 97 after different neutron irradiation conditions

Directory of Open Access Journals (Sweden)

Christian Dethloff

2016-12-01

Full Text Available Characterization of irradiation induced microstructural evolution 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 determine the defect microstructure after different neutron irradiation conditions. In particular dislocation loops, voids and precipitates are analyzed concerning defect nature, density and size distribution after irradiation to 15 dpa at 300 °C in the mixed spectrum High Flux Reactor (HFR. New results are combined with previously obtained data from irradiation in the fast spectrum BOR-60 reactor (15 and 32 dpa, 330 °C, which allows for assessment of dose and dose rate effects on the aforementioned irradiation induced defects and microstructural characteristics.

Irradiation effects on the ductility of fusion reactor structural materials

International Nuclear Information System (INIS)

Boudamous, F.

1986-10-01

Austenitic and ferritic-martensitic stainless steels have been proposed as first wall structural materials for the next generation of fusion devices. In order to study the effect of high temperature irradiation on their tensile properties, specimens of the steel AISI 316 L (CEC reference), of the martensitic steel W. Nr 1.4914 and of the duplex ferritic-martensitic steel EM12 have been irradiated in the BR2 reactor in Mol. The austenitic steel was irradiated at 470 0 C to about 1.1 10 22 n/cm 2 ( E>0.1 MeV) while the ferritic-martensitic steels were irradiated at 590 0 C to about 7.7 10 22 n/cm 2 (E>0.1 MeV). The tensile tests of the 316 L steel have been performed between 250 and 750 0 C. Below around 550 0 C, the yield stress after irradiation increased from about 160 to 270 MPa and the total elongation decreased from 42 to about 26%. At 750 0 C, the yield stress increase was small but the total elongation decreased from 60 to only 10%. At this temperature, the rupture of the irradiated specimen was intergranular while all the other specimens presented a transgranular rupture. At 650 0 C the variations were intermediate. The change of the ultimate tensile strength was small at all test temperatures. The EM12 and W. Nr 1.4914 steels tested only at 550 0 C, showed a decrease of the yield and tensile strength as well as an increase of the total elongation. The same tests performed on specimens which have been heat treated in parallel showed that the observed changes were due, in a large part, if not completely, to the maintenance of steels at high temperature

Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

DEFF Research Database (Denmark)

Singh, B.N; Edwards, D.J.; Bilde-Sørensen, Jørgen

2004-01-01

The phenomenon of plastic flow localization in the form of "cleared" channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels areinitiated during post-irradiation deformation...... has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons.Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron...... irradiated at 323 and 373K to displacement doses in the range of 0.01 to 0.3 dpa (displacement per atom) and tensile tested at the irradiation temperature.The stress-strain curves clearly indicated the occurrence of a yield drop. The post-deformation microstructural examinations revealed that the channels...

Tensile Properties of Under-Matched Weld Joints for 950 MPa Steel.

Science.gov (United States)

Yamamoto, Kouji; Arakawa, Toshiaki; Akazawa, Nobuki; Yamamoto, Kousei; Matsuo, Hiroki; Nakagara, Kiyoyuki; Suita, Yoshikazu

In welding of 950 MPa-class high tensile strength steel, preheating is crucial in order to avoid cold cracks, which, however, eventually increases welding deformations. One way to decrease welding deformations is lowering preheating temperature by using under-matched weld metal. Toyota and others clarify that although breaking elongation can decrease due to plastic constraint effect under certain conditions, static tensile of under-matched weld joints is comparable to that of base metal. However, there has still been no report about joint static tensile of under-matched weld joints applied to 950 MPa-class high tensile strength steel. In this study, we aim to research tensile strength and fatigue strength of under-matched weld joints applied to 950 MPa-class high tensile steel.

Evaluation of the mechanical properties of carbon fiber after electron beam irradiation

International Nuclear Information System (INIS)

Giovedi, Claudia; Diva Brocardo Machado, Luci; Augusto, Marcos; Segura Pino, Eddy; Radino, Patricia

2005-01-01

Carbon fibers are used as reinforcement material in epoxy matrix in advanced composites. An important aspect of the mechanical properties of composites is associated to the adhesion between the surface of the carbon fiber and the epoxy matrix. This paper aimed to the evaluation of the effects of EB irradiation on the tensile properties of two different carbon fibers prepared as resin-impregnated specimens. The fibers were EB irradiated before the preparation of the resin-impregnated specimens for mechanical tests. Observations of the specimens after breakage have shown that EB irradiation promoted significant changes in the failure mode. Furthermore, the tensile strength data obtained for resin-impregnated specimens prepared with carbons fibers previously irradiated presented a slight tendency to be higher than those obtained from non-irradiated carbon fibers

Pre- and post-irradiation properties of copper alloys at 250 deg. C following bonding and bakeout thermal cycles

International Nuclear Information System (INIS)

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

1997-01-01

Screening experiments were carried out to investigate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing, and bonding thermal treatment followed by re-ageing and the reactor bakeout treatment at 350 deg. C for 100 h. Tensile specimens of CuAl-25 were given the heat treatment corresponding to the bonding thermal cycle. A number of heat treated specimens were neuron irradiated at 250 deg. C to a dose level of ∼ 0.3 dpa in the DR-3 reactor at Risoe. Both unirradiated and irradiated specimens with various heat treatments were tensile tested at 250 deg. C. The microstructure and electrical resistivity of these specimens were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure of the irradiated specimens was also investigated. The fracture surfaces of both unirradiated and irradiated specimens were examined. Results of these investigations are reported in the present report. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250 deg. C showed a severe loss of ductility in the case of CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens, on the other hand, exhibited a reasonable amount of uniform elongation. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure. (au) 7 tabs., 28 ills., 15 refs

Study of I11-conditioning of Linac stereotactic irradiation subspaces using singular values decomposition analysis

International Nuclear Information System (INIS)

Platoni, K.; Lefkopoulos, D.; Grandjean, P.; Schlienger, M.

1999-01-01

A Linac sterotactic irradiation space is characterized by different angular separations of beams because of the geometry of the stereotactic irradiation. The regions of the stereotactic space characterized by low angular separations are one of the causes of ill-conditioning of the stereotactic irradiation inverse problem. The singular value decomposition (SVD) is a powerful mathematical analysis that permits the measurement of the ill-conditioning of the stereotactic irradiation problem. This study examines the ill-conditioning of the stereotactic irradiation space, provoked by the different angular separations of beams, using the SVD analysis. We subdivided the maximum irradiation space (MIS: (AA) AP x (AA) RL =180 x 180 ) into irradiation subspaces (ISSs), each characterized by its own angular separation. We studied the influence of ISSs on the SVD analysis and the evolution of the reconstruction quality of well defined three-dimensional dose matrices in each configuration. The more the ISS is characterized by low angular separation the more the condition number and the reconstruction inaccuracy are increased. Based on the above results we created two reduced irradiation spaces (RIS: (AA) AP x (AA) RL =180 x 140 and (AA) AP x (AA) RL =180 x 120 ) and compared the reconstruction quality of the RISs with respect to the MIS. The more an irradiation space is free of low angular separations the more the irradiation space contains useful singular components. (orig.)

In-vitro Degradation Behaviour of Irradiated Bacterial Cellulose Membrane

International Nuclear Information System (INIS)

Darwis, D.; Khusniya, T.; Hardiningsih, L.; Nurlidar, F.; Winarno, H.

2012-01-01

Bacterial cellulose membrane synthesized by Acetobacter xylinum in coconut water medium has potential application for Guided bone Regeneration. However, this membrane may not meet some application requirements due to its low biodegradation properties. In this paper, incorporation of gamma irradiation into the membrane is a developed strategy to increase its biodegradability properties. The in-vitro degradation study in synthetic body fluid (SBF) of the irradiated membrane has been analyzed during periods of 6 months by means of weight loss, mechanical properties and scanning electron microscopy observation compared to that the un-irradiated one. The result showed that weight loss of irradiated membrane with 25 kGy and 50 kGy and immersed in SBF solution for 6 months reached 18% and 25% respectively. While un-irradiated membrane did not give significant weight loss. Tensile strength of membranes decreases with increasing of irradiation dose and further decreases in tensile strength is observed when irradiated membrane was followed by immersion in SBF solution. Microscope electron image of cellulose membranes shows that un-irradiated bacterial cellulose membrane consists of dense ultrafine fibril network structures, while irradiation result in cleavage of fibrils network of cellulose. The fibrils network become loosely after irradiated membrane immersed in SBF solution due to released of small molecular weight carbohydrates formed during by irradiation from the structure (author)

Effect of 200 keV proton irradiation on the properties of methyl silicone rubber

International Nuclear Information System (INIS)

Zhang Lixin; Xu Zhou; Wei Qiang; He Shiyu

2006-01-01

The effects of 200 keV proton irradiation on methyl silicone rubber were studied. The changes in surface morphology, mechanical properties, cross-linking density, glass transition temperature, infrared attenuated total reflection spectrum and mass spectrum indicated that, at lower fluence, the proton irradiation induced cross-linking, resulting in an increase in tensile strength and hardness of the methyl silicone rubber. However, at higher proton fluence, radiation-induced degradation, which decreased the tensile strength and hardness, became dominant. A macromolecular-network destruction model for silicone rubber irradiated with protons was proposed

Effects of tensile and compressive stresses on irradiation-induced swelling in AISI 316

International Nuclear Information System (INIS)

Lauritzen, T.; Bell, W.L.; Konze, G.M.; Rosa, J.M.; Vaidyanathan, S.; Garner, F.A.

1985-05-01

The results of two recent experiments indicate that the current perception of stress-affected swelling needs revision. It appears that compressive stresses do not delay swelling as previously modeled but actually accelerate swelling at a rate comparable to that induced by tensile stresses

Post irradiation examination of RAF/M steels after fast reactor irradiation up to 33 dpa and < 340 C (ARBOR1). RAFM steels. Metallurgical and mechanical characterisation. Final report for TW2-TTMS-001b, D9

Energy Technology Data Exchange (ETDEWEB)

Petersen, C. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). EURATOM, Inst. fuer Materialforschung, Programm Kernfusion

2010-07-01

In an energy generating fusion reactor structural materials will be exposed to very high dpa-levels of about 100 dpa. Due to this fact and because fast reactor irradiation facilities in Europe are not available anymore, a reactor irradiation at the State Scientific Center of the Russian Federation with its Research Institute of Atomic Reactors (SSC RIAR), Dimitrovgrad, had been performed in the fast reactor BOR 60 with an instrumented test rig. This test rig contained tensile, impact and Low Cycle Fatigue type specimens used at FZK since many years. Samples of actual Reduced Activation Ferritic/Martensitic (RAF/M) -steels (e.g. EUROFER 97) had been irradiated in this reactor at a lower temperature (< 340 C) up to a damage of 33 dpa. This irradiation campaign was called ARBOR 1. Starting in 2003 one half of these irradiated samples were post irradiation examined (PIE) by tensile testing, low cycle fatigue testing and impact testing under the ISTC Partner Contract 2781p in the hot cells of SSC RIAR. In the post irradiation instrumented impact tests a significant increase in the Ductile to Brittle Transition Temperature as an effect of irradiation has been detected. During tensile testing the strength values are increasing and the strain values reduced due to substantial irradiation hardening. The hardening rate is decreasing with increasing damage level, but it does not show saturation. The low cycle fatigue behaviour of all examined RAF/M - steels show at total strain amplitudes below 1 % an increase of number of cycles to failure, due to irradiation hardening. From these post irradiation experiments, like tensile, low cycle fatigue and impact tests, radiation induced design data, e.g. for verification of design codes, can be generated.

A study of the tensile behaviour of flax tows and their potential for composite processing

International Nuclear Information System (INIS)

Moothoo, J.; Allaoui, S.; Ouagne, P.; Soulat, D.

2014-01-01

Highlights: • Flax tows are characterised by tensile testing under various conditions. • The tensile properties and their dispersion are dependent on the gauge length. • The distribution of the fibres length in specimens commands the tensile behaviour. • Packs of fibre bundle debonding failure mode are observed by image correlation. • Interesting tensile properties are obtained on epoxy impregnated and cured tows. - Abstract: To study the potential of flax tows in composite processing as an alternative to flax spun yarns, a flat flax tow consisting of aligned fibre bundles held together by a natural binder was used and characterised in tension under various conditions. The effect of the gauge length was studied on the dry reinforcement. The experimental results showed that the mechanical properties and failure mechanism varied significantly depending on the gauge length and are discussed in relation to the distribution of elementary fibres within the tow. A characteristic length as from which the mechanical properties are stable has been identified. At this length, the effect of the strain rate on the tensile properties was measured and their sensitivity to the strain rate suggests a viscous effect in the behaviour of the flax tow. To approach process conditions such as wet filament winding, a batch of specimens was impregnated with epoxy prior to tensile testing. The tensile properties under wet conditions were found to be close to the properties under dry conditions and shows that the tow can withstand typical processing tensions. Finally, tensile tests on cured-impregnated tows showed interesting mechanical properties for composite application

Molecular Dynamics Simulations of Tensile Behavior of Copper

OpenAIRE

Sainath, G.; Srinivasan, V. S.; Choudhary, B. K.; Mathew, M. D.; Jayakumar, T.

2014-01-01

Molecular dynamics simulations on tensile deformation of initially defect free single crystal copper nanowire oriented in {100} has been carried out at 10 K under adiabatic and isothermal loading conditions. The tensile behaviour was characterized by sharp rise in stress in elastic regime followed by sudden drop at the point of dislocation nucleation. The important finding is that the variation in dislocation density is correlated with the observed stress-strain response. Several interesting ...

Hemostimulating efficiency of non-steroid anti-inflammatory drugs under modified irradiation conditions

International Nuclear Information System (INIS)

Zhvoronkov, L.P.; Sklobovskaya, I.Eh.

1988-01-01

Non-steroid anti-inflammatory drugs (NSAID) were found to have hemostimulating effect in mice after irradiation. This effect was rather definite under irradiation conditions modified by dose fractioning or radioprotective chemicals. NSAID application during fractionated irradiation with midlethal integral dose leads to almost complete recovery of bone marrow hemopoiesis by the 9th day of radiation illness. NSAID usage combined with chemical radioprotectors provides effective hemopoiesis stimulation leading to survival increase in animals, irradiated with absolutely lethal doses. (author)

Kenaf Fibre Reinforced Polypropylene Composites: Effect of Cyclic Immersion on Tensile Properties

Directory of Open Access Journals (Sweden)

W. H. Haniffah

2015-01-01

Full Text Available This research studied the degradation of tensile properties of kenaf fibre reinforced polypropylene composites due to cyclic immersion into two different solutions, as well as comparison of the developed composites’ tensile properties under continuous and cyclic immersion. Composites with 40% and 60% fibre loadings were immersed in tap water and bleach for 4 cycles. Each cycle consisted of 3 days of immersion and 4 days of conditioning in room temperature (28°C and 55% humidity. The tensile strength and modulus of composites were affected by fibre composition, type of liquid of immersion, and number of cycles. The number of immersion cycles and conditioning caused degradation to tensile strength and modulus of kenaf fibre reinforced polypropylene composites. Continuous and cyclic immersion in bleach caused tensile strength of the composites to differ significantly whereas, for tensile modulus, the difference was insignificant in any immersion and fibre loadings. However, continuous immersion in the bleach reduced the tensile strength of composites more compared to cyclic immersion. These preliminary results suggest further evaluation of the suitability of kenaf fibre reinforced polypropylene composites for potential bathroom application where the composites will be exposed to water/liquid in cyclic manner due to discontinuous usage of bathroom.

Characterization of mechanical properties and microstructure of highly irradiated SS 316

Energy Technology Data Exchange (ETDEWEB)

Karthik, V., E-mail: karthik@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kumar, RanVijay; Vijayaragavan, A.; Venkiteswaran, C.N.; Anandaraj, V.; Parameswaran, P.; Saroja, S.; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.; Jayakumar, T.; Raj, Baldev [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2013-08-15

Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673–823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

International Nuclear Information System (INIS)

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

1995-01-01

Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5x10 21 to 1.5x10 24 n/m 2 (E>1 MeV) with a flux of 2.5x10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of similar 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ∼ =0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources. ((orig.))

The Effect of Gamma Irradiation on the Mechanical Properties of vulcanized Natural Rubber and Natural Rubber-Polyethylene Blends

International Nuclear Information System (INIS)

Sudradjat Iskandar

2008-01-01

To enhance the quality of vulcanized natural rubber and natural rubber-polyethylene blends, gamma irradiation has been done. The compound of natural rubber and natural rubber-polyethylene blends made by using roll mill machine. The mixed materials were antioxidant, anti ozon, plasticizer and vulcanisator. The natural rubber and natural rubber-polyethylene blends compound were vulcanizer and made a slab (film of sample) using hot and could press machine. The slabs produced were then gamma irradiated at irradiation dose of 75, 150 and 300 kGy. Before and after irradiation, the slab were characterized using strograph R1 machine. The results showed that the modulus 300 and hardness of vulcanized natural rubber and natural rubber-polyethylene blends were increasing; the tensile strength and tear strength were increasing to maximum level then decreasing with gamma irradiation, while the elongation at break was decreasing. The maximum tensile strength of vulcanized natural rubber and natural rubber-polyethylene blends were found at irradiation dose of 75 kGy. At the irradiation dose of 75 kGy, the tensile strength of vulcanized natural rubber increased from 17.6 MN/m 2 to 21.2 MN/m 2 , while the tensile strength of vulcanized natural rubber-polyethylene blends increased slightly from 18.7 MN/m 2 to 19.4 MN/m 2 . (author)

Solar irradiance forecasting at one-minute intervals for different sky conditions using sky camera images

International Nuclear Information System (INIS)

Alonso-Montesinos, J.; Batlles, F.J.; Portillo, C.

2015-01-01

Highlights: • The solar resource has been predicted for three hours at 1-min intervals. • Digital image levels and cloud motion vectors are joint for irradiance forecasting. • The three radiation components have been predicted under different sky conditions. • Diffuse and global radiation has an nRMSE value around 10% in all sky conditions. • Beam irradiance is predicted with an nRMSE value of about 15% in overcast skies. - Abstract: In the search for new techniques to predict atmospheric features that might be useful to solar power plant operators, we have carried out solar irradiance forecasting using emerging sky camera technology. Digital image levels are converted into irradiances and then the maximum cross-correlation method is applied to obtain future predictions. This methodology is a step forward in the study of the solar resource, essential to solar plant operators in adapting a plant’s operating procedures to atmospheric conditions and to improve electricity generation. The results are set out using different statistical parameters, in which beam, diffuse and global irradiances give a constant normalized root-mean-square error value over the time interval for all sky conditions. The average measure is 25.44% for beam irradiance; 11.60% for diffuse irradiance and 11.17% for global irradiance.

Tensile properties of the modified 13Cr martensitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Mabruri, Efendi, E-mail: effe004@lipi.go.id; Anwar, Moch Syaiful, E-mail: moch.syaiful.anwar@lipi.go.id; Prifiharni, Siska, E-mail: siska.prifiharni@lipi.go.id; Romijarso, Toni B.; Adjiantoro, Bintang [Research Center for Metallurgy and Materials, Indonesian Institute of Sciences (LIPI) Kawasan Puspiptek Gd. 470 Serpong, Tangerang Selatan 15314 (Indonesia)

2016-04-19

This paper reports the influence of Mo and Ni on the tensile properties of the modified 13Cr martensitic stainless steels in tempered condition. Four steels with different content of Mo and Ni were prepared by induction melting followed by hot forging, quenching and tempering. The experimental results showed that the addition of about 1% and 3% Mo has a beneficial effect to increase both the tensile strength and the elongation of the steels. On the contrary, the addition of about 3% Ni into the martensitic stainless steel results in decreasing of both the tensile strength and the elongation. Among the alloys investigated the 13Cr3Mo type steel exhibited largest tensile strength of 1348 MPa and largest elongation of 12%. The observation on the tensile fractured surfaces by using scanning electron microscope supported these findings.

Tensile properties of the modified 13Cr martensitic stainless steels

International Nuclear Information System (INIS)

Mabruri, Efendi; Anwar, Moch Syaiful; Prifiharni, Siska; Romijarso, Toni B.; Adjiantoro, Bintang

2016-01-01

This paper reports the influence of Mo and Ni on the tensile properties of the modified 13Cr martensitic stainless steels in tempered condition. Four steels with different content of Mo and Ni were prepared by induction melting followed by hot forging, quenching and tempering. The experimental results showed that the addition of about 1% and 3% Mo has a beneficial effect to increase both the tensile strength and the elongation of the steels. On the contrary, the addition of about 3% Ni into the martensitic stainless steel results in decreasing of both the tensile strength and the elongation. Among the alloys investigated the 13Cr3Mo type steel exhibited largest tensile strength of 1348 MPa and largest elongation of 12%. The observation on the tensile fractured surfaces by using scanning electron microscope supported these findings.

Tensile Bond Strength of Metal Bracket Bonding to Glazed Ceramic Surfaces With Different Surface Conditionings

Directory of Open Access Journals (Sweden)

M. Imani

2011-12-01

Full Text Available Objective: The objective of this study was to compare the tensile bond strength of metal brackets bonding to glazed ceramic surfaces using three various surface treatments.Materials and Methods: Forty two glazed ceramic disks were assigned to three groups. In the first and second groups the specimens were etched with 9.5% hydrofluoric acid (HFA. Subsequently in first group, ceramic primer and adhesive were applied, but in second group a bonding agent alone was used. In third group, specimens were treated with 35% phosphoric acid followed by ceramic primerand adhesive application. Brackets were bonded with light cure composites. The specimens were stored in distilled water in the room temperature for 24 hours and thermocycled 500 times between 5°C and 55°C. The universal testing machine was used to test the tensile bond strength and the adhesive remenant index scores between three groups was evaluated. The data were subjected to one-way ANOVA, Tukey and Kruskal-Wallis tests respectively.Results: The tensile bond strength was 3.69±0.52 MPa forfirst group, 2.69±0.91 MPa for second group and 3.60±0.41 MPa for third group. Group II specimens showed tensile strength values significantly different from other groups (P<0.01.Conclusion: In spite of limitations in laboratory studies it may be concluded that in application of Scotch bond multipurpose plus adhesive, phosphoric acid can be used instead of HFA for bonding brackets to the glazed ceramic restorations with enough tensile bond strength.

Final report on graphite irradiation test OG-2

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1975-01-01

Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P 3 JHAN were also included

HVEM in-situ deformation of neutron irradiated Fe-0.3a/oCu

International Nuclear Information System (INIS)

Gelles, D.S.; Thomas, L.E.; Powell, R.W.

1980-01-01

In an effort to better understand the nature of irradiation embrittlement of low alloy steels, in-situ HVEM deformation tests have been performed on pure iron and iron-0.3% copper ribbon tensile specimens in both the unirradiated and irradiated (2.5 x 10 19 n/cm 2 , E > 1 MeV at 290 0 C) conditions. Microstructural response is described principally in terms of the matrix in which cracks were observed to propagate across the different specimens. Major differences between the irradiation iron-copper alloy and the other material conditions were (1) the plastic zone which developed ahead to propagating cracks was smaller, and (2) dislocations were found to develop a crenulated structure uncharacteristic of the other conditions studied. It is inferred that the crenulations result from the presence of obstacles to dislocation motion and thereby demonstrate the matrix strengthening effect of small, radiation-induced, presumably copper-rich precipitates. However, the obstacles which are most effective in retarding dislocation motion are found to be distributed on a much coarser scale than that of the radiation-induced loops

Oil palm empty fruit bunch (OPEFB) fiber reinforced PVC/ENR blend-electron beam irradiation

International Nuclear Information System (INIS)

Ratnam, Chantara Thevy; Raju, Gunasunderi; Wan Md Zin Wan Yunus

2007-01-01

The effect of irradiation on the tensile properties of oil palm empty fruit bunch (OPEFB) fiber reinforced poly(vinyl chloride)/epoxidized natural rubber (PVC/ENR) blends were studied. The composites were prepared by mixing the fiber and the PVC/ENR blend using HAAKE Rheomixer at 150 deg. C. The composites were then irradiated by using a 3.0 MeV electron beam machine at doses ranging from 0 to 100 kGy in air and room temperature. The tensile strength, Young's modulus, elongation at break and gel fraction of the composites were measured. Comparative studies were also made by using poly(methyl acrylate) grafted OPEFB fiber in the similar blend system. An increase in tensile strength, Young's modulus and gel fraction, with a concurrent reduction in the elongation at break (Eb) of the PVC/ENR/OPEFB composites were observed upon electron beam irradiation. Studies revealed that grafting of the OPEFB fiber with methyl acrylate did not cause appreciable effect to the tensile properties and gel fraction of the composites upon irradiation. The morphology of fractured surfaces of the composites, examined by a scanning electron microscope showed an improvement in the adhesion between the fiber and the matrix was achieved upon grafting of the fiber with methyl acrylate

The effect of deformation twinning on irradiation embrittlement in iron single crystals

International Nuclear Information System (INIS)

Kayano, Hideo; Tokutomi, Shoichiro; Yajima, Seishi; Takaku, Hiroshi.

1978-01-01

Single crystals of iron with the [100] crystal orientation were irradiated in JMTR with fast neutrons to a fluence of 8 x 10 18 n/cm 2 (E > 1 MeV). All samples were deformed in tension at temperatures from liquid nitrogen temperature to 200 0 C at different strain rates using an Instron-type tensile testing machine. Scanning electron microscopy of the fractured surfaces revealed that deformation twinning is difficult to occur in irradiated samples, and also that twins formed in both irradiated and unirradiated samples inhibit fracture nucleation and growth. From the results of tensile deformation of the irradiated samples deformed in tension a different strain rates at 159 0 K, it is conceived that twinning suppression is greater in the irradiated than in the unirradiated samples, and that the nucleation and growth of twins are not necessarily related to those of cracks. It is suggested that the irradiation-induced defects impede plastic deformation of the crystals and deformation twinning is suppressed by irradiation, thus causing the irradiation embrittlement. (auth.)

Investigation of some physical properties of polypropylene irradiated by gamma rays

International Nuclear Information System (INIS)

Kattn, M.; Ajji, Z.

2005-03-01

Pure polypropylene samples were exposed to different of gamma radiation up to 100 kGy in presence of oxygen or nitrogen. Some physical properties were investigated in relation to the radiation dose: melting point, crystallinity, apparent activation energy; tensile strength,; elongation. The data show that the crystallinity decreases at low doses. In addition, the melting point is shifted to lower temperature with increasing the irradiation dose. The apparent activation energy increases with increasing irradiation dose. The tensile strength increases for low doses up to maximum, and after this value it decreases increasing (Authors)

Effects of gamma Rays Irradiation on resistance of Pseudomonas aeruginosa in various condition

International Nuclear Information System (INIS)

Nikham

2002-01-01

The investigation of gamma tays 60Co irradiation effect on resistance of bacteri P.aeruginosa has been done.The objective of the research was to know the D10 value of bacteria P.aeruginosa. By using of distilled water,talc and peanut powder as carrier in dry,wet,O 2 and N 2 condition the bacteria of P.aeruginosa were irradiated on gamma rays of 6 0Co with dose of O to 2.5 kGy,and with dose rate of 5 and 10 kGy/h.After irradiation the bacteria of P. aeruginosa were cultured in media of the Tryptone Soya Agar and incubatedat temperature of 32±2 o C for 3 days. The survival colonies were calculated,and the data were used to make the curve and to determine the D10 value. The results of the experiments showed that D10 value of irradiated bacteria of P.aeruginosain the disitilled water,talc and peanut powder as carrier were not high significant.Nevertheless the D10 value of the irradiated at dose rate 10kGy/h show more higher tendency than at dose rate 5kGy/h. The D10 value of irradiated bacteria in the N2 condition was higher,if compared with in the O2 condition

Final report on neutron irradiation at low temperature to investigate plastic instability and a high temperature to study cavitation

International Nuclear Information System (INIS)

Singh, B.N.; Eldrup, M.; Golubov, S.I.; Edwards, D.J.; Jung, P.

2005-01-01

Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at 333 and 573 K. Electrical resistivity and tensile properties were measured both in the unirradiated and irradiated condition. Some additional specimens of pure iron were irradiated at 333 K to doses of 10-3, 10-2 and 10-1 dpa and tensile tested at 333 K. To investigate the effect of helium on cavity nucleation and growth, specimens of pure iron and EUROFER 97 were implanted with different amounts of helium at 323 K and subsequently neutron irradiated to doses of 10-3, 10-2 and 10-1 dpa at 323 K. Defect microstructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations, based on the production bias model (PBM) were carried out to study the details of evolution of cavities with and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been studied using the Kinetic Monte Carlo (KMC) code during a realistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops.The present report describes both experimental procedure and calculational methodology employed in the present work. The main results of all these investigations, both experimental and theoretical, are highlighted with appropriate examples. Finally, a brief summary is given of the main results conclusions. (au)

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Grounes, M [AB Atomenergi, Nykoeping (Sweden); Myers, H P [Chalmers Institute of Technology, Goeteborg (Sweden); Hannerz, N E [Motala Verkstads AB, Motala (Sweden)

1967-09-15

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10{sup 18} and 5.6 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature.

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

International Nuclear Information System (INIS)

Grounes, M.; Myers, H.P.; Hannerz, N.E.

1967-09-01

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10 18 and 5.6 x 10 18 n/cm 2 (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature

Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

International Nuclear Information System (INIS)

McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

1999-01-01

Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

Oxygen sensitization of mammalian cells under different irradiation conditions

International Nuclear Information System (INIS)

Ling, C.C.; Michaels, H.B.; Gerweck, L.E.; Epp, E.R.; Peterson, E.C.

1981-01-01

The oxygen dependence of the radiosensitivity of cultured CHO cells was examined in detail with particular attention paid to avoiding possible artifacts due to radiolytic oxygen depletion. Two methods of gas equilibration and irradiation were used. In the first approach, cells were irradiated with 50-kVp X rays in a thin-layer geometry which offered maximum interchange between the cells and the surrounding gas. The second technique employed 280-kVp X irradiation of cells under full-medium conditions with mechanical agitation to minimize the effect of radiochemical oxygen consumption by promoting rapid oxygen replenishment. With these techniques oxygen radiosensitization was clearly resolved at an oxygen concentration of 0.03% in the gas phase. The oxygen K curves measured by these two methods were similar in shape over a wide range of oxygen concentration

Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

Energy Technology Data Exchange (ETDEWEB)

Alsabbagh, Ahmad, E-mail: ahalsabb@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Sarkar, Apu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Miller, Brandon [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Burns, Jatuporn [Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Squires, Leah; Porter, Douglas; Cole, James I. [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Murty, K.L. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

2014-10-06

Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) have 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 1.37 dpa. Atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. Mechanical properties were characterized using microhardness and tensile tests, and irradiation of UFG carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. After irradiation, micro hardness indicated increases of around 9% for UFG versus 62% for CG steel. Similarly, tensile strength revealed increases of 8% and 94% respectively for UFG and CG steels while corresponding decreases in ductility were 56% versus 82%. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between experimental results and modeling were demonstrated based on irradiation induced precipitate strengthening and dislocation forest hardening mechanisms.

Isotopic tailoring with 59Ni to study the effect of helium on microstructural evolution and mechanical properties of neutron-irradiated Fe-Cr-Ni alloys

International Nuclear Information System (INIS)

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

1992-03-01

Tensile testing on three model Fe-Cr-Ni alloys removed from four discharges of the 59 Ni isotopic doping experiment in FFTF-MOTA indicates that helium/dpa ratios typical of fusion reactors do not produce changes in the yield strength or elongation that are significantly different from those at much lower helium generation rates. It also appears that tensile properties approach a saturation level that is dependent only on the final irradiation temperature, but not prior temperature history or thermomechanical starting condition. The saturation in mechanical properties reflects a similar saturation in microstructure that is independent of starting condition. The successful conduct of an isotopic doping experiment was found to require post-irradiation measurement of the helium levels in order to compensate for uncertainties in the cross sections for burn-out and burn-in of 59 Ni and for uncertainties in neutron flux and spectra in the vicinity of the edge of the core

High-Tensile Strength Tape Versus High-Tensile Strength Suture: A Biomechanical Study.

Science.gov (United States)

Gnandt, Ryan J; Smith, Jennifer L; Nguyen-Ta, Kim; McDonald, Lucas; LeClere, Lance E

2016-02-01

To determine which suture design, high-tensile strength tape or high-tensile strength suture, performed better at securing human tissue across 4 selected suture techniques commonly used in tendinous repair, by comparing the total load at failure measured during a fixed-rate longitudinal single load to failure using a biomechanical testing machine. Matched sets of tendon specimens with bony attachments were dissected from 15 human cadaveric lower extremities in a manner allowing for direct comparison testing. With the use of selected techniques (simple Mason-Allen in the patellar tendon specimens, whip stitch in the quadriceps tendon specimens, and Krackow stitch in the Achilles tendon specimens), 1 sample of each set was sutured with a 2-mm braided, nonabsorbable, high-tensile strength tape and the other with a No. 2 braided, nonabsorbable, high-tensile strength suture. A total of 120 specimens were tested. Each model was loaded to failure at a fixed longitudinal traction rate of 100 mm/min. The maximum load and failure method were recorded. In the whip stitch and the Krackow-stitch models, the high-tensile strength tape had a significantly greater mean load at failure with a difference of 181 N (P = .001) and 94 N (P = .015) respectively. No significant difference was found in the Mason-Allen and simple stitch models. Pull-through remained the most common method of failure at an overall rate of 56.7% (suture = 55%; tape = 58.3%). In biomechanical testing during a single load to failure, high-tensile strength tape performs more favorably than high-tensile strength suture, with a greater mean load to failure, in both the whip- and Krackow-stitch models. Although suture pull-through remains the most common method of failure, high-tensile strength tape requires a significantly greater load to pull-through in a whip-stitch and Krakow-stitch model. The biomechanical data obtained in the current study indicates that high-tensile strength tape may provide better repair

Influence of Immersion Conditions on The Tensile Strength of Recycled Kevlar®/Polyester/Low-Melting-Point Polyester Nonwoven Geotextiles through Applying Statistical Analyses

Directory of Open Access Journals (Sweden)

Jing-Chzi Hsieh

2016-05-01

Full Text Available The recycled Kevlar®/polyester/low-melting-point polyester (recycled Kevlar®/PET/LPET nonwoven geotextiles are immersed in neutral, strong acid, and strong alkali solutions, respectively, at different temperatures for four months. Their tensile strength is then tested according to various immersion periods at various temperatures, in order to determine their durability to chemicals. For the purpose of analyzing the possible factors that influence mechanical properties of geotextiles under diverse environmental conditions, the experimental results and statistical analyses are incorporated in this study. Therefore, influences of the content of recycled Kevlar® fibers, implementation of thermal treatment, and immersion periods on the tensile strength of recycled Kevlar®/PET/LPET nonwoven geotextiles are examined, after which their influential levels are statistically determined by performing multiple regression analyses. According to the results, the tensile strength of nonwoven geotextiles can be enhanced by adding recycled Kevlar® fibers and thermal treatment.

Tensile bond strength of metal bracket bonding to glazed ceramic surfaces with different surface conditionings.

Science.gov (United States)

Akhoundi, Ms Ahmad; Kamel, M Rahmati; Hashemi, Sh Mahmood; Imani, M

2011-01-01

The objective of this study was to compare the tensile bond strength of metal brackets bonding to glazed ceramic surfaces using three various surface treatments. Forty two glazed ceramic disks were assigned to three groups. In the first and second groups the specimens were etched with 9.5% hydrofluoric acid (HFA). Subsequently in first group, ceramic primer and adhesive were applied, but in second group a bonding agent alone was used. In third group, specimens were treated with 35% phosphoric acid followed by ceramic primer and adhesive application. Brackets were bonded with light cure composites. The specimens were stored in distilled water in the room temperature for 24 hours and thermocycled 500 times between 5°C and 55°C. The universal testing machine was used to test the tensile bond strength and the adhesive remenant index scores between three groups was evaluated. The data were subjected to one-way ANOVA, Tukey and Kruskal-Wallis tests respectively. The tensile bond strength was 3.69±0.52 MPa forfirst group, 2.69±0.91 MPa for second group and 3.60±0.41 MPa for third group. Group II specimens showed tensile strength values significantly different from other groups (Ptensile bond strength.

Damage-induced tensile instability

International Nuclear Information System (INIS)

Hult, J.

1975-01-01

The paper presents a unified description of ductile and brittle rupture phenomena in structural components under tensile loading with particular emphasis on creep rupture. Two structural elements are analyzed in detail: 1) the uniform tensile bar subject to a Heaviside history of tensile force and superimposed such loadings, i.e. staircase histories, and 2) the thinwalled spherical pressure vessel subject to a Heaviside history of internal pressure. For both these structures the conditions for instantaneous as well as delayed rupture are analysed. It is shown that a state of mechanical instability will be reached at a certain load or after a certain time. The cases of purely ductile rupture and purely brittle fracture are identified as two limiting cases of this general instability phenomenon. The Kachanov-Rabotnov damage law implies that a structural component will fail in tension only when it has reached a state of complete damage, i.e. zero load carrying capacity. The extended law predicts failure at an earlier stage of the deterioration process and is therefore more compatible with experimental observation. Further experimental support is offered by predictions for staircase loading histories, both step-up and step-down type. The presented damage theory here predicts strain histories which are in closer agreement with test data than predictions based on other phenomenological theories

SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement

International Nuclear Information System (INIS)

Van Walle, E.; Chaouadi, R.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

1998-01-01

The contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV weld material is reported. The objective of this contribution is twofold: (1) to gain experience in the field of the testing of WWER-440 steels; (2) to analyse the round-robin data according to in-house developed on used models in order to check their validity and applicability. Results from testing on unirradiated material are reported including data obtained from chemical analysis, Charpy-V impact testing, tensile testing and fracture toughness determination. Finally, irradiation strategies that can be used in the program to obtain irradiated, irradiated-annealed and irradiated-annealed-reirradiated conditions are outlined

Evaluation of ferritic alloy Fe-2 1/4Cr-1Mo after neutron irradiation: Microstructural development

International Nuclear Information System (INIS)

Gelles, D.S.

1986-10-01

As part of a program to provide a data base on the bainitic alloy Fe-2-1/4-1Mo for fusion energy applications, microstructural examinations are reported for nine specimen conditions for 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510 0 C. Void swelling is found following irradiation at 400 0 C to 480 0 C. Concurrently dislocation structure and precipitation developed. Peak void swelling, void density, dislocation density and precipitate number density formed at the lowest temperature, approximately 400 0 C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior

Effects of neutron irradiation and fatigue on ductility of stainless steel DIN 1.4948

International Nuclear Information System (INIS)

Vries, M.I. de; Schaaf, B. van der; Staal, H.U.; Elen, J.D.

1978-10-01

Test specimens of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 neutrons (n).m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High temperature (723 K to 1023 K) tensile tests at strain rates (depsilon/dt) from 10 -7 s -1 to 10 s -1 show a considerable decrease of tensile ductility. The extent depends on helium content, test temperature and strain rate. The atomic helium fractions of 3.10 -7 and 7.10 -6 result from the reactions of thermal neutrons with the 14 ppm boron, present in the steel. Helium embrittlement sets in at strain rates below 1 s -1 to 10 s -1 (the range of interest for Bethe-Tait accident analyses). A minimum total elongation value of 6% is shown at 923 K. The post-irradiation fatigue life is reduced by up to about 50% due to intergranular cracking. The combination of irradiation and fatigue causes a decrease of ductility after a smaller number of prior fatigue cycles than in the case of unirradiated material. (Auth.)

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

Influence of LBE long term exposure and simultaneous fast neutron irradiation on the mechanical properties of T91 and 316L

Energy Technology Data Exchange (ETDEWEB)

Stergar, E., E-mail: estergar@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Eremin, S.G. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation); Gavrilov, S.; Lambrecht, M. [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Makarov, O.; Iakovlev, V. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

2016-05-15

The LEXUR–II–LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

Study of the effect of gamma irradiation on carbon black loaded low-density polyethylene films

International Nuclear Information System (INIS)

Salem, M.A.; Hussein, A.; El-Ahdal, M.A.

2003-01-01

The effect of gamma irradiation on the tensile and physico-chemical properties of low-density polyethylene (LDPE) films loaded with different concentrations of carbon black (C.B) has been studied. The results showed that the behavior of the samples during gamma irradiation is complicated and this may be due to scission and the interaction between oxidation and crosslinking processes. The tensile properties are modified by the presence of carbon black. Film sample containing 7% C.B was found to exhibit a nearly stabilized tensile behavior with radiation dose, which allows to use this formulation in packaging for food sterilization and in preservation of weak cobalt-gamma sources. (author)

Conditions of viscosity measurement for detecting irradiated peppers

International Nuclear Information System (INIS)

Hayashi, Toru; Todoriki, Setsuko; Okadome, Hiroshi; Kohyama, Kaoru

1995-01-01

Viscosity of gelatinized suspensions of black and white peppers decreased depending upon dose. The viscosity was influenced by gelatinization and viscosity measurement conditions. The difference between unirradiated pepper and an irradiated one was larger at a higher pH and temperature for gelatinization. A viscosity parameter normalized with the starch content of pepper sample and the viscosity of a 5% suspension of corn starch could get rid of the influence of the conditions for viscosity measurement such as type of viscometer, shear rate and temperature. (author)

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

International Nuclear Information System (INIS)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B.

2015-01-01

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer.

Biosorption of the strontium ion by irradiated Saccharomyces cerevisiae under culture conditions.

Science.gov (United States)

Qiu, Liang; Feng, Jundong; Dai, Yaodong; Chang, Shuquan

2017-06-01

As a new-emerging method for strontium disposal, biosorption has shown advantages such as high sorption capacity; low cost. In this study, we investigated the potential of Saccharomyces cerevisiae (S. cerevisiae) in strontium disposal under culture conditions and the effects of irradiation on their biosorption capabilities. We found that S. cerevisiae can survive irradiation and grow. Pre-exposure to irradiation rendered S. cerevisiae resistant to further irradiation. Surprisingly, the pre-exposure to irradiation can increase the biosorption capability of S. cerevisiae. We further investigated the factors that influenced the biosorption efficiency, which were (strongest to weakest): pH > strontium concentration > time > temperature. In our orthogonal experiment, the optimal conditions for strontium biosorption by irradiated S. cerevisiae were: pH 7, 150 mg L -1 strontium at the temperature of 32 °C with 30 h. The equilibrium of strontium biosorption was analyzed by Langmuir and Freundlich models, from which the formal model is found to provide a better fit for the experimental results. The kinetics of strontium biosorption by living irradiated S. cerevisiae was found to be comprised of three phases: dramatically increased during 0-9 h, decreased during 12-24 h, and increased during 30-50 h. These results provide a systematic understanding of the biosorption capabilities of irradiated S. cerevisiae, which can contribute to the development of remediating nuclear waste water. Copyright © 2017 Elsevier Ltd. All rights reserved.

Mechanical properties of freeze-dried and irradiated bone chips, fascia lata and dura mater

International Nuclear Information System (INIS)

De Guzman, Z.M.; Vajaradul, Y.

1996-01-01

The comparison strengths of freeze-dried and irradiated bone chips such as three-dimensional cortex (3DC) and two-cortico cancellous (2CC) are investigated. The results show that the (3DC) exhibits a higher compression strength (1.2kN cm -2 in deep frozen states. Rehydration of the freeze-dried bone chips after 15 min with normal saline solution restores the strength of materials by 30%. The tensile strengths of fascia lata and dura mater are also studied. A marked decrease of tensile strength is noted in the irradiated and freeze-dried samples, however, reconstitution with normal saline solution restores the tensile strength of the tissues to about 40-56%. (author). 8 refs., 6figs

Effect of irradiation on the properties of some shrinking polymer films

International Nuclear Information System (INIS)

Varsanyi, E.

1974-01-01

Shrinking polymer films (polyethylene, polyvinylidene chloride, polyester) suitable for use in the food industry were studied with the intention to determine the effect of radurizing doses (800 krad and below) on changes in the proportion of crystalline parts in the polymer, and on the tensile strength, elongation at break and shrinkage of the film. Changes in the crystalline/amorphous ratio in the polymer were determined by means of infra-red spectrophotometry. Calculations based on spectral data showed no significant changes in the ratio of crystalline fraction of any of the films, as a function of radurizing doses. Tensile strength and elongation at break tests were carried out by means of standardized instruments and methods. It was found that the tensile strength of the polyethylene film decreased by about 25% as an effect of irradiation, while the same treatment caused no significant changes in the elongation at break. The tensile strength of the polyvinylidene chloride film suffered a decrease of roughly 15%, its elongation at break an about 30% decrease when irradiated. Radiation treatment caused a decrease if less than 10% in tensile strength of the polyester film and a more than 10% change in elongation at break. The tests indicated no significant changes in the shrinkage of radiation treated polymers. The results of the tests led to the conclusion that radurizing doses caused no such change which would affect the applicability of polymer films to the wrapping and packaging of foods subjected to irradiation or would make the films unsuitable for the protection of the goods. (F.J.)

Study and simulation of irradiated zirconium alloys fracture under type RIA accidental loading conditions; Comprehension et modelisation de la rupture d'alliages de zirconium irradies en conditions accidentelles de type RIA

Energy Technology Data Exchange (ETDEWEB)

Le Saux, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

2008-07-01

The thesis aims to study and simulate the mechanical behavior under Reactivity Initiated Accident loading conditions, of the Zircaloy 4 fuel claddings, irradiated or not. It also aims to characterize and simulate the behavior and the fracture under RIA loading conditions of hydrided Zircaloy 4 non irradiated. This study proposes an experimental approach and a simulation. (A.L.B.)

Modification of the properties of NBR/EPDM blends vulcanized by gamma irradiation

International Nuclear Information System (INIS)

Abou Zeid, M.M.; Shaltout, N.A.; Mohamed, M.A.; El Miligy, A.A.

2001-01-01

Blends of nitrile-butadiene rubber, NBR with ethylene propylene diene monomer EPDM rubber with varying contents have been prepared. Unloaded or loaded blends with 40 phr of HAF carbon black have been vulcanized by using gamma irradiation. Mechanical properties, namely tensile strength, tensile modulus and elongation at break have been followed up as a function of irradiation dose as well as blend component compositions. Moreover, the susceptibility of prepared composites towards organic solvents and car oils has been followed up in terms of swelling number and soluble fraction measurements. The organic solvents used are toluene and dimethyl-formamide and oil are car lubricating and brake oils. The results indicated improvements in mechanical properties of blend composites with irradiation dose and increased content NBR in the blend. Also, susceptibility to fluids decreased appreciably with irradiation dose but with different extents for different fluids

Effect of helium and DPA's on tensile properties of V-5Ti and V-3Ti-1Si

International Nuclear Information System (INIS)

Witzenburg, W. van; Vries, E. de.

1991-02-01

Specimens of the alloys V-5Ti and V-3Ti-1Si were irradiated in a mixed-spectrum fission reactor in reactor grade liquid sodium to a fast neutron fluence of 3.8 x 10 25 m -2 (E>0.1 MeV), which corresponds to 6.2 dpa. Irradiation temperatures were 500, 600 and 700 deg C. Some of the specimens were pre-injected with helium to 100 appm at approx 50 deg C by means of a cyclotron. In addition, part of the specimens were doped with boron-10 to concentrations of 100 and 600 appm. Tensile testing, at temperatures equal to the irradiation temperatures and at a strain rate of 10 -4 s -1 , showed an increase in strength and reduced elongation at 500 deg C and to a lesser extent at 600 deg C. These changes are caused by displacement damage. Helium, pre-injected as well as produced by transmutation of boron-10, did not have a significant influence on the tensile properties. Cavities seen in the irradiated materials at low concentrations, were not preferentially located on grain boundaries. There was no apparent deleterious effect of lithium, which is also a transmutation product of boron-10. (author). 12 refs.; 8 figs.; 3 tabs

Tensile strength and fatigue strength of 6061 aluminum alloy friction welded joints

Energy Technology Data Exchange (ETDEWEB)

Ochi, H.; Tsujino, R. [Osaka Inst. of Tech., Asahi-ku Osaka (Japan); Sawai, T. [Osaka Sangyo Univ., Daito (Japan); Yamamoto, Y. [Setsunan Univ., Neyagawa (Japan); Ogawa, K. [Osaka Prefecture Univ., Sakai (Japan); Suga, Y. [Keio Univ., Kohoku-ku, Yokohama (Japan)

2002-07-01

Friction welding of 6061 aluminum alloy was carried out in order to examine the relationship between deformation heat input in the upset stage and joint performance. The joint performance was evaluated by tensile testing and fatigue testing. Stabilized tensile strength was obtained when the deformation heat input in the upset stage exceeded 200 J/s. Weld condition at the weld interface and the width of softened area affected fatigue strength more than tensile strength. That is, when the weld condition at the weld interface is good and the softened area is wide, fatigue strength increases. On the other hand, when the weld condition at the weld interface is good and the softened area is narrow, and when the weld condition at the weld interface is somewhat poor in spite of the wide softened area, fatigue strength decreases. The fatigue limit obtained by the fatigue testing revealed that, when the deformation heat input in the upset stage exceeded a certain value, sound joints could be produced. (orig.)

Tensile behaviour and properties of a bone analogue composite (HA, HDPE) crosslinked by gamma radiation

International Nuclear Information System (INIS)

Romero, G.; Smolko, Eduardo E.

2005-01-01

A natural composite material, hydroxyapatite (HA) and high density polyethylene (HDPE) crosslinked by ionizing radiations is been developed as a bioactive analogue material for bone replacement. Mechanical properties of the composites irradiated up to 300 kGy under tensile tests was studied. Gel content and micrographs of different composite fractures are shown. (author)

Effects of electron beam irradiation on ethylene-octene copolymers (octene rubber)

International Nuclear Information System (INIS)

Harris C Raj Kumar; Mansor Ahmad; Khairul Zaman Mohd Dahlan

2002-01-01

The effect of electron irradiation on a ethylene-octene copolymer was investigated. The optimal blending speed, blending temperature and hot press temperature were first optimized to 40 rpm, 185 degree C and 180 degree C, respectively. The ethylene octene copolymer was then irradiated with electron beam from doses in the range of 20 kGy up to 200 kGy. The physical changes occurred were examined from the point of tensile strength tests, elongation at break, tensile modulus, hardness (Shore A) and gel content, and compared with a set of un-irradiated sample. Almost all the tests signify that cross-linking was the predominant reaction rather than chain scission, especially in gel content test. The hardness test was inconclusive as there were no significant changes that occurred. (Author)

Comparison of properties and microstructures of Trefimetaux CuNiBe and Hycon 3HP TM before and after neutron irradiation. (ITER R and D Task no. T213)

International Nuclear Information System (INIS)

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

2000-07-01

The precipitation strengthened CuNiBe alloys are among the three candidate copper alloys that are 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 Cu-A1 2 O 3 alloys. Brush-Wellman Inc. has developed an improved version of their Hycon CuNiBe alloy that has higher conductivity while maintaining a reasonable level of strength. In the present work we have investigated the physical and mechanical properties of the Hycon 3HP TM alloy both before and after neutron irradiation and have compared its microstructure and properties with the European CuNiBe candidate alloy manufactured by Trefimetaux. Tensile specimens of both alloys were irradiated in the DR-3 reactor at Risoe to displacement dose levels of up to 0.3 dpa at 100, 250 and 350 d eg C . Both alloys were tensile tested in vacuum in the unirradiated and irradiated conditions at 100, 250 and 350 d eg C and the microstructures of the alloys were investigated using transmission electron microscopy. Electrical resistivity measurements were made on tensile specimens be-fore and after irradiation; all measurements were made at 23 d eg C . Results of these investigations are presented and discussed in terms of the sensitivity of these alloys to test temperature, which becomes increasingly problematic when the irradiation and test temperature reaches 250 d eg C and above. (au)

Assessment of repair welding technologies of irradiated materials

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Damages of reactor internals of stainless steels caused by SCC and fatigue were identified in aged BWR plants. Repair-welding is one of the practical countermeasure candidates to restore the soundness of components and structures. The project of 'Assessment of Repair welding Technologies of Irradiated Materials' is being carried out to develop the technical guideline regarding the repair-welding of reactor internals. In fiscal 2011, we investigated the weldability of stainless steel 316L irradiated by welding (TIG) tungsten inert gas. Furthermore, the tensile properties and stress corrosion cracking (SCC) susceptibility of the welds were investigated. Cross-sectional observation of heat affected zone (HAZ) of the bead on plate TIG weldments (heat input 4 kJ/cm) of irradiated SUS316L stainless steel containing 0.026 ~ 0.12appm helium showed degradation of grain boundaries due to helium accumulation. Degree of the degradation depended on the amount of helium. No deterioration of grain boundaries was observed by bead on plate welding with one pass one layer when helium content was 0.039appm. The tensile strengths of welds in non-irradiated and irradiated material were similar. However, the elongation of a weldment by irradiated SUS316L containing 0.124appm Helium was lower than non-irradiated. It was estimated to cause the effects of helium bubbles. The SCC susceptibility of the HAZ was no significant difference compared with other locations. (author)

Research and application of fuzzy subtractive clustering model on tensile strength of radiation vulcanization for nitrile-butadiene rubber

International Nuclear Information System (INIS)

Zuo Duwen; Wang Hong; Zhu Nankang

2010-01-01

By use of fuzzy subtractive clustering model, the relationship between tensile strength of radiation vulcanization of NBRL (Nitrile-butadiene rubber latex) and irradiation parameters have been investigated. The correlation coefficient was calculated to be 0.8222 in the comparison of experimental data to the predicted data. It was obvious that fuzzy model identification method is not only high precision with small computation, but also easy to be used. It can directly supply the evolution of tensile strength of NBR by fuzzy modeling method in radiation vulcanization process for nitrile-butadiene rubber. (authors)

Instrumented indentation for characterization of irradiated metals at room and high temperatures

International Nuclear Information System (INIS)

Sacksteder, Irene

2011-01-01

The reliability and sustainability of future fusion power plants will highly depend on the aptitude of materials to withstand severe irradiation conditions induced by the burning plasma in reactors. The so-called reduced-activation ferritic-martensitic (RAFM) steels are the current promising candidates for the structural applications considering the reactor's first wall. These steels exhibit irradiation embrittlement and hardening for defined irradiation conditions that are mainly characterized by the irradiation temperature and the irradiation dose. A proper characterization of such irradiated steels implies the use of adapted mechanical testing tools. In the present study, the instrumented indentation technique makes use of a post-processing tool based on neural networks. This technique has been selected for its ability to examine tensile properties by multistage indents on miniaturized irradiated metallic samples. The steel specimens studied in this project have been neutron-irradiated up to a dose of 15 dpa. They have been subsequently tested at room temperature in a Hot Cell by means of an adapted commercial indentation device. The significant irradiation-induced hardening effect present in the range of 250-350 deg C could be observed in the hardness and material's strength parameters. These two material parameters show a similar evolution with increasing irradiation temperatures. Post-irradiation annealing treatments of Eurofer97 have been realized and leads to a partial recovery of the irradiation damage. Considering the demands for characterization in irradiated steels at high temperature and for post-irradiation annealing experiments, the existing instrumented indentation device has been further developed during this work. A conceptual design has been proposed for an indentation testing machine, operating at up to 650 deg C, while remaining the critical temperature limit for tensile strength of the newly developed oxide dispersion strengthening ferritic

Effect of Short PET Fiber and Electron Beam Irradiation on The Properties of Acrylonitrile Butadiene Rubber-Poly(Vinyl Chloride) (NBR-PVC) Blend

International Nuclear Information System (INIS)

Youssef, H.A.; Shaltout, N.A.; EI Nemer, K.F.; EI Miligy, A.A.

2009-01-01

Blend of acrylonitrile-butadiene rubber (NBR ) and ploy vinyl chloride(PYV) (70-30) has been loaded with different concentrations of polyethylene terephthalate (PET) fibers waste ( 0.5 - 40 p hr); in the presence of resorcinol hexamethylenetetramine - precipitated silica (RHS) as bonding agent system and pentaeritheroal tetraacrylate (PET A) as co agent. Curing of the prepared composites has been carried out by electron beam irradiation (25 - 150 kGy) under atmospheric conditions. Evaluations of mechanical, physical, and thermal properties of uncured as well as cured composites have been undertaken. It has been found that the tensile strength, tensile modulus at 25 % elongation and hardness were increased with irradiation dose as well as fiber loading whereas the elongation at break and soluble fraction were decreased. Moreover, it has been found that the thermal stability of prepared composites at constant fiber loading of 10 p hr is improved on irradiation up to 100 kGy. Confirmation of latter data has been found through calculation of activation energy, Ea of the thermal degradation process

Effect of Short PET Fiber and Electron Beam Irradiation on The Properties of Acrylonitrile Butadiene Rubber-Poly(Vinyl Chloride) (NBR-PVC) Blend

International Nuclear Information System (INIS)

Youssef, H.A.; Shaltout, N.A.; EI Nemer, K.F.; EI Miligy, A.A.

2008-01-01

Blend of acrylonitrile-butadiene rubber (NBR ) and ploy vinyl chloride(PYV) (70-30) has been loaded with different concentrations of polyethylene terephthalate (PET) fibers waste ( 0.5 - 40 p hr); in the presence of resorcinol hexamethylenetetramine - precipitated silica (RHS) as bonding agent system and pentaeritheroal tetraacrylate (PET A) as co agent. Curing of the prepared composites has been carried out by electron beam irradiation (25 - 150 kGy) under atmospheric conditions. Evaluations of mechanical, physical, and thermal properties of uncured as well as cured composites have been undertaken. It has been found that the tensile strength, tensile modulus at 25 % elongation and hardness were increased with irradiation dose as well as fiber loading whereas the elongation at break and soluble fraction were decreased. Moreover, it has been found that the thermal stability of prepared composites at constant fiber loading of 10 p hr is improved on irradiation up to 100 kGy. Confirmation of latter data has been found through calculation of activation energy, Ea of the thermal degradation process

Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

International Nuclear Information System (INIS)

Griger, Agnes

2004-01-01

The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

Effect of tensile stress on cavitation damage formation in mercury

Energy Technology Data Exchange (ETDEWEB)

Naoe, Takashi, E-mail: naoe.takashi@jaea.go.j [J-PARC Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kogawa, Hiroyuki [J-PARC Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Yamaguchi, Yoshihito [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Futakawa, Masatoshi [J-PARC Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

2010-03-15

Cavitation erosion or so called pitting damage was investigated under tensile stress conditions in mercury. In MW-class liquid metal spallation targets, pitting damage is a critical issue to satisfy required power and/or lifetime of the target vessel. Cavitation occurs by negative pressure which is induced through pressure wave propagation due to proton beam injection. Pitting damage is formed by microjet and/or shock wave during cavitation bubble collapse. A mercury target vessel suffers tensile stress due to thermal stress or welding. In order to investigate the effect of tensile stress on pitting damage formation, cavitation erosion tests were performed using stress imposed specimens in mercury. An ultrasonic vibratory horn and electro-Magnetic IMpact Testing Machine (MIMTM) were used to vary the cavitation intensity. In the incubation period of pitting damage, damaged area was slightly increased with increasing imposed tensile stress. In the steady state period, a mean depth of erosion was increased by the tensile stress. Additionally, in order to quantitatively evaluate the effect of tensile stress, an indentation test with Vickers indenter was carried out to quasi-statically simulate the impact load. From the measurement of the diagonal length of the indent aspect ratio and hardness, it is recognized that the threshold of the deformation, i.e. pitting damage formation, was decreased by the tensile stress.

The effect of irradiation on the mechanical properties of 6061-T651 aluminum

International Nuclear Information System (INIS)

Alexander, D.J.

1992-01-01

Critical components of the Advanced Neutron Source (ANS) reactor, to be built at Oak Ridge National Laboratory (ORNL), will be fabricated from 6061-T651 aluminum alloy. This alloy has been selected for its favorable neutronic, thermal, and mechanical properties. The effect of irradiation on the tensile properties and fracture toughness has been studied to allow the lifetime of these components to be estimated. Irradiations were carried out in the High Flux Isotope Reactor at ORNL at a temperature of approximately 95 degree C to a fluence of approximately 10 26 m -2 (thermal). Testing was conducted from room temperature to 150 degree C. The yield and ultimate tensile strengths were increased by irradiation, and the total elongation decreased, but the fracture toughness at 26 and 95 degree C was not degraded by irradiation, and decreased only slightly at 150 degree C

Dynamics of a stabilized motor defense conditioned reflex at different levels of motivation in irradiated rats

Energy Technology Data Exchange (ETDEWEB)

Shtemberg, A S

1982-05-01

Postradiation dynamics of strengthened motor-defense conditioned reflex in rats-males irradiated with the doses of 94.111 and 137 Gy was studied. Phase disturbances of conditioned-reflex activity increased with enhancing irradiation dose have been revealed. Rapid recovery of conditioned reflex after short primary aggravation was a characteristic peculiarity. At that, the dynamics of relation of main nervous processes in cortex was noted for significant instability increasing with radiation syndrome development. Enhancement of force of electro-defense support promoted more effective strengthening of temporary connections and conditioned high stability of trained-reflex reactions during serious functional disturbances resulted from sublethal dose irradiation.

Effect of helium irradiation on fracture modes

International Nuclear Information System (INIS)

Hanamura, T.; Jesser, W.A.

1982-01-01

The objective of this work is to determine the crack opening mode during in-situ HVEM tensile testing and how it is influenced by test temperature and helium irradiation. Most cracks were mixed mode I and II. However, between 250 0 C and room temperature the effect of helium irradiation is to increase the amount of mode I crack propagation. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in irradiated specimens tested between 250 0 C and room temperature, but could be restored by a post irradiation anneal

Radiation deterioration of several aromatic polymers under oxidative conditions

International Nuclear Information System (INIS)

Sasuga, Tsuneo; Hagiwara, Miyuki

1987-01-01

Radiation-induced oxidative irradiation effects (with γ-rays under oxygen pressure) or poly(aryl sulphones) (U-PS and PES), poly(aryl ester) (U-Polymer), poly(aryl amide) (A-Film) and poly(aryl ether ether ketone) (PEEK) have been studied based on changes in tensile properties. The deterioration dose estimated from the decrease in the elongation at break was as low as one-fifth to one-tenth of that in high-dose-rate electron-beam irradiation, but the order of radiation resistance of the polymers did not differ from that in electron-beam irradiation, i.e. PEEK > A-Film > U-Polymer > U-PS > PES. The radiation stability of aromatic units under oxidative conditions was estimated from a comparison of the radiation resistance of the polymers themselves and their chemical structures. The following order was obtained: diphenyl ether, diphenyl ketone > aromatic amide>> bisphenol A > diphenyl sulphone. The deterioration mechanism of PEEK under oxidative irradiation was studied by measuring dynamic viscoelastic properties. It was concluded that deterioration in mechanical properties under oxidative irradiation was brought about by chain scission only. (author)

Irradiation Creep in Graphite

Energy Technology Data Exchange (ETDEWEB)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

Effect of Microwave Disinfection on Compressive and Tensile Strengths of Dental Stones

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Mahmood Robati Anaraki

2013-03-01

Full Text Available Background and aims. Although microwave irradiation has been used for disinfection of dental stone casts, there are concerns regarding mechanical damage to casts during the process. The aim of this study was to evaluate the effect of microwave irradiation on the compressive strength (CS and diametral tensile strength (DTS of stone casts. Materials and methods. In this in vitro study, 80 cylindrical type III and IV stone models (20 × 40 mm were prepared and divided into 8 groups of 10. The DTS and CS of the specimens were measured by a mechanical testing machine at a crosshead speed of 0.5 cm/min after 7 times of frequent wetting, irradiating at an energy level of 600 W for 3 minutes and cooling. Data were analyzed by Student’s t-test. Results. Microwave irradiation significantly increased DTS of type III and IV to 5.23 ± 0.64 and 8.17 ± 0.94, respectively (P < 0.01. Conclusion. According to the results, microwave disinfection increases DTS of type III and IV stone casts without any effects on their CS.

Evaluation and Study the Effect of Additives and Other Factors on Tensile Strength of Asphalt Paving Mixtures

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Hanaa Khaleel A. Al-Baiti

2012-03-01

Full Text Available The resistance of asphaltic concrete to cracking is dependent upon its tensile strength and flexibility characteristics. Also the low tensile strength has recognized as a major contributor to other performance problems. The fatigue life of mixtures decreases exponentially with decreasing of tensile strength. This trend is justified by the loss in stiffness and thereby initiating cracks and stripping. The main objective of this research is intended to study the effect of different variables related with the used materials and the external conditions on the tensile strength and predict a model of indirect tensile strength in asphalt concrete paving materials under the local prevailing conditions and investigate the effect of percent of additives of (Polyestrene resins and Hydrated Lime to enhance the resistance ability of asphalt concrete mixture against distresses. The main affected factors; soaking, asphalt content, compaction, aggregate maximum size and temperature, influence on the indirect tensile strength and presented through a statistics analysis model for tensile strength in asphalt mixture

Behavior of pre-irradiated fuel under a simulated RIA condition

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

1994-07-01

This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Miya, Naoyuki

2002-01-01

In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

Wood Sawdust/Natural Rubber Ecocomposites Cross-Linked by Electron Beam Irradiation

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Elena Manaila

2016-06-01

Full Text Available The obtaining and characterization of some polymeric eco-composites based on wood sawdust and natural rubber is presented. The natural rubber was cross-linked using the electron beam irradiation. The irradiation doses were of 75, 150, 300 and 600 kGy and the concentrations of wood sawdust were of 10 and 20 phr, respectively. As a result of wood sawdust adding, the physical and mechanical properties such as hardness, modulus at 100% elongation and tensile strength, showed significant improvements. The presence of wood sawdust fibers has a reinforcing effect on natural rubber, similar or better than of mineral fillers. An increase in the irradiation dose leads to the increasing of cross-link density, which is reflected in the improvement of hardness, modulus at 100% elongation and tensile strength of blends. The cross-linking rates, appreciated using the Flory-Rehner equation, have increased with the amount of wood sawdust in blends and with the irradiation dose. Even if the gel fraction values have varied irregularly with the amount of wood sawdust and irradiation dose it was over 90% for all blends, except for the samples without wood sawdust irradiated with 75 kGy. The water uptake increased with increasing of fiber content and decreased with the irradiation dose.

Effect of implanted helium on tensile properties and hardness of 9% Cr martensitic stainless steels

Science.gov (United States)

Jung, P.; Henry, J.; Chen, J.; Brachet, J.-C.

2003-05-01

Hundred micrometer thick specimens of 9% Cr martensitic steels EM10 and T91 were homogeneously implanted with He 4 to concentrations up to 0.5 at.% at temperatures from 150 to 550 °C. The specimens were tensile tested at room temperature and at the respective implantation temperatures. Subsequently the fracture surfaces were analysed by scanning electron microscopy and some of the specimens were examined in an instrumented hardness tester. The implanted helium caused hardening and embrittlement which both increased with increasing helium content and with decreasing implantation temperature. Fracture surfaces showed intergranular brittle appearance with virtually no necking at the highest implantation doses, when implanted below 250 °C. The present tensile results can be scaled to tensile data after irradiation in spallation sources on the basis of helium content but not on displacement damage. An interpretation of this finding by microstructural examination is given in a companion paper [J. Nucl. Mater., these Proceedings].

Mechanical behaviour of neutron irradiated Nb monocrystalline

International Nuclear Information System (INIS)

Otero, M.P.; Lucki, G.

1986-01-01

Nb [941] - oriented single crystal was irradiated to a fluence of 1,1 x 10 19 n/cm 2 in the IEA-R1 reactor at IPEN-CNEN/SP. Tensile-Stress experiments showed an irradiation induced hardening, characterized by an increase in the yield stress of about 16%. This result was interpreted using the 'lattice hardening' model. The observed slip systems are attributed to the gliding of the anomalous slip planes. (Author) [pt

Radiation resistance of polymer materials. Degradation evaluation by accelerated testing for application condition

International Nuclear Information System (INIS)

Seguchi, Tadao; Tamura, Kiyotoshi; Sorimachi, Masami

2010-02-01

This paper presents re-evaluated radiation resistance property data of polymer materials, which had been tested in past times in TAKASAKI Quantum Beam Science Directorate, for the future study of ageing evaluation of low voltage electric cable insulation materials used in light-water nuclear reactors. The radiation resistance of 25 types of plastics and rubbers materials applied in practical environments was evaluated by the accelerated testing of gamma-ray irradiation under oxygen pressure, and was compared with the radiation resistance determined from the traditional testing by irradiation with a high dose rate in air. The polymer materials were formulated to be similar or equivalent to practical materials, and the most of formulation (chemical compounds and quantities) were described. For all materials, the tensile properties (elongation at break, ultimate strength, 100% or 200% modulus), electric resistivity, gel-fraction, and density were measured after irradiation in oxidation conditions and irradiation in air with a high dose rate (non-oxidation conditions). The data of relations between each properties and total dose at various conditions were compiled, and the relations among the changes of mechanical properties, electrical properties, and radiation induced chemical reactions were discussed. (author)

Effect of parenteral nutrition on the bone marrow recovery under exsperimental conditions of uneven irradiation

International Nuclear Information System (INIS)

Moroz, B.B.; Fedorovskij, L.L.; Deshevoj, Yu.B.

1986-01-01

Using white rats-males the effect of parenteral nutrition (PN) on blood formation (hemopoiesis) recovery under the conditions of total (control), subtotal (shielding of animal hind limb) with 7.5 Gy and X-ray partial irradiation of abdomen region with the 13.5 Gy has been studied. It has been found that bone marrow recovery increases either at subtotal or partial irradiation under the conditions of PN. Mechanisms of PN favourable effect are discussed. The advisability of using PN under uneven irradiations of organism is indicated

Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

International Nuclear Information System (INIS)

Zinkle, S.J.

1998-01-01

The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of ∼5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule

Relaxation behavior of radicals produced in irradiated black pepper under various moisture conditions by ESR

International Nuclear Information System (INIS)

Kameya, Hiromi; Kawauchi, Risa; Shimoyama, Yuhei; Kaeda, Yoko; Ogawa, Satoko; Nakamura, Hideo; Ukai, Mitsuko

2008-01-01

Black pepper is easy to be contaminated by microorganism and often processed to γ-irradiation. ESR has been used for the detection of radicals induced in irradiated spices. Using ESR, we revealed the effects of moisture condition during storage of irradiated black pepper on the saturation behavior of ESR signal. The ESR spectrum of black pepper consists of a broad sextet centered at g=2.0, a singlet as same g-value and a singlet at g=4.0. The irradiation causes two new signals, one is the strong and sharp singlet signal at g=2.0 and the other is the side signal. We found that the signal intensity originated by the radicals of black pepper with and without radiation decayed in the high humidity condition during storage. The ESR signal intensity of irradiated black pepper decayed during storage and showed almost the same intensity level as that of non-irradiated black pepper during storage. (author)

Effect of ion irradiation on the surface, structural and mechanical properties of brass

Energy Technology Data Exchange (ETDEWEB)

Ahmad, Shahbaz; Bashir, Shazia, E-mail: shaziabashir@gcu.edu.pk; Ali, Nisar; Umm-i-Kalsoom,; Yousaf, Daniel; Faizan-ul-Haq,; Naeem, Athar; Ahmad, Riaz; Khlaeeq-ur-Rahman, M.

2014-04-01

Highlights: • Brass targets were exposed to carbon ions of energy 2 MeV. • The effect of ion dose has been investigated. • The surface morphology is investigated by SEM analysis. • XRD analysis is performed to reveal structural modification. • Mechanical properties were investigated by tensile testing and microhardness testing. - Abstract: Modifications to the surface, structural and mechanical properties of brass after ion irradiation have been investigated. Brass targets were bombarded by carbon ions of 2 MeV energy from a Pelletron linear accelerator for various fluences ranging from 56 × 10{sup 12} to 26 × 10{sup 13} ions/cm{sup 2}. A scanning electron microscope and X-ray diffractometer were utilized to analyze the surface morphology and crystallographic structure respectively. To explore the mechanical properties e.g., yield stress, ultimate tensile strength and microhardness of irradiated brass, an universal tensile testing machine and Vickers microhardness tester were used. Scanning electron microscopy results revealed an irregular and randomly distributed sputter morphology for a lower ion fluence. With increasing ion fluence, the incoherently shaped structures were transformed into dendritic structures. Nano/micro sized craters and voids, along with the appearance of pits, were observed at the maximum ion fluence. From X-ray diffraction results, no new phases were observed to be formed in the brass upon irradiation. However, a change in the peak intensity and higher and lower angle shifting were observed, which represents the generation of ion-induced defects and stresses. Analyses confirmed modifications in the mechanical properties of irradiated brass. The yield stress, ultimate tensile strength and hardness initially decreased and then increased with increasing ion fluence. The changes in the mechanical properties of irradiated brass are well correlated with surface and crystallographic modifications and are attributed to the generation

Laser solder welding of articular cartilage: tensile strength and chondrocyte viability.

Science.gov (United States)

Züger, B J; Ott, B; Mainil-Varlet, P; Schaffner, T; Clémence, J F; Weber, H P; Frenz, M

2001-01-01

The surgical treatment of full-thickness cartilage defects in the knee joint remains a therapeutic challenge. Recently, new techniques for articular cartilage transplantation, such as mosaicplasty, have become available for cartilage repair. The long-term success of these techniques, however, depends not only on the chondrocyte viability but also on a lateral integration of the implant. The goal of this study was to evaluate the feasibility of cartilage welding by using albumin solder that was dye-enhanced to allow coagulation with 808-nm laser diode irradiation. Conventional histology of light microscopy was compared with a viability staining to precisely determine the extent of thermal damage after laser welding. Indocyanine green (ICG) enhanced albumin solder (25% albumin, 0.5% HA, 0.1% ICG) was used for articular cartilage welding. For coagulation, the solder was irradiated through the cartilage implant by 808-nm laser light and the tensile strength of the weld was measured. Viability staining revealed a thermal damage of typically 500 m in depth at an irradiance of approximately 10 W/cm(2) for 8 seconds, whereas conventional histologies showed only half of the extent found by the viability test. Heat-bath investigations revealed a threshold temperature of minimum 54 degrees C for thermal damage of chondrocytes. Efficient cartilage bonding was obtained by using bovine albumin solder as adhesive. Maximum tensile strength of more than 10 N/cm(2) was achieved. Viability tests revealed that the thermal damage is much greater (up to twice) than expected after light microscopic characterization. This study shows the feasibility to strongly laser weld cartilage on cartilage by use of a dye-enhanced albumin solder. Possibilities to reduce the range of damage are suggested. Copyright 2001 Wiley-Liss, Inc.

Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

International Nuclear Information System (INIS)

Singh, B.N.; Xiaoxu Huang; Taehtinen, S.; Moilamen, P.; Jacquet, P.; Dekeyser, J.

2007-11-01

Traditionally, the effect of irradiation on mechanical properties of metals and alloys is determined using post-irradiation tests carried out on pre-irradiated specimens and in the absence of irradiation environment. The results of these tests may not be representative of deformation behaviour of materials used in the structural components of a fission or fusion reactor where the materials will be exposed concurrently to displacement damage and external and/or internal stresses. In an effort to evaluate and understand the dynamic response of materials under these conditions, we have recently performed a series of uniaxial tensile tests on Fe-Cr and pure iron specimens in the BR-2 reactor at Mol (Belgium). The present report first provides a brief description of the test facilities and the procedure used for performing the in-reactor tests. The results on the mechanical response of materials during these tests are presented in the form of stress-displacement dose and the conventional stress-strain curves. For comparison, the results of post-irradiation tests and tests carried out on unirradiated specimens are also presented. Results of microstructural investigations on the unirradiated and deformed, irradiated and undeformed, post-irradiation deformed and the in-reactor deformed specimens are also described. During the in-reactor tests the specimens of both Fe-Cr alloy and pure iron deform in a homogeneous manner and do not exhibit the phenomenon of yield drop. An increase in the pre-yield dose increases the yield stress but not the level of maximum flow stress during the in-reactor deformation of Fe-Cr alloy. Neither the in-reactor nor the post-irradiation deformed specimens of Fe-Cr alloy and pure iron showed any evidence of cleared channel formation. Both in Fe-Cr and pure iron, the in-reactor deformation leads to accumulation of dislocations in a homogeneous fashion and only to a modest density. No dislocation cells are formed during the in-reactor or post-irradiation

Effect of ion irradiation on the surface, structural and mechanical properties of brass

Science.gov (United States)

Ahmad, Shahbaz; Bashir, Shazia; Ali, Nisar; Umm-i-Kalsoom; Yousaf, Daniel; Faizan-ul-Haq; Naeem, Athar; Ahmad, Riaz; Khlaeeq-ur-Rahman, M.

2014-04-01

Modifications to the surface, structural and mechanical properties of brass after ion irradiation have been investigated. Brass targets were bombarded by carbon ions of 2 MeV energy from a Pelletron linear accelerator for various fluences ranging from 56 × 1012 to 26 × 1013 ions/cm2. A scanning electron microscope and X-ray diffractometer were utilized to analyze the surface morphology and crystallographic structure respectively. To explore the mechanical properties e.g., yield stress, ultimate tensile strength and microhardness of irradiated brass, an universal tensile testing machine and Vickers microhardness tester were used. Scanning electron microscopy results revealed an irregular and randomly distributed sputter morphology for a lower ion fluence. With increasing ion fluence, the incoherently shaped structures were transformed into dendritic structures. Nano/micro sized craters and voids, along with the appearance of pits, were observed at the maximum ion fluence. From X-ray diffraction results, no new phases were observed to be formed in the brass upon irradiation. However, a change in the peak intensity and higher and lower angle shifting were observed, which represents the generation of ion-induced defects and stresses. Analyses confirmed modifications in the mechanical properties of irradiated brass. The yield stress, ultimate tensile strength and hardness initially decreased and then increased with increasing ion fluence. The changes in the mechanical properties of irradiated brass are well correlated with surface and crystallographic modifications and are attributed to the generation, augmentation, recombination and annihilation of the ion-induced defects.

Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys

International Nuclear Information System (INIS)

Matijasevic, Milena; Al Mazouzi, Abderrahim

2008-01-01

High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems (ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement. In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160 deg. C to 300 deg. C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300 deg. C. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening. (authors)

Comparative study of the tungsten irradiation conditions in IFMIF and DEMO

Energy Technology Data Exchange (ETDEWEB)

Simakov, S.P.; Pereslavtsev, P.; Fischer, U. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Moeslang, A. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Material Research I

2010-05-15

The International Fusion Material Irradiation Facility (IFMIF) [1] will provide an accelerator based intense neutron source with a white spectrum extending up to 55 MeV for high fluence irradiations of fusion power reactor (FPR) candidate materials. Material samples located in test modules will be subjected to a radiation load anticipated for a fusion power reactor. The highest neutron flux is expected in the High Flux Test Module, which is considered in the IFMIF design to host around 1000 compactly packed stainless steel samples - the main structure materials of power fusion reactors. Another material subjected to the highest loads in a FPR is a tungsten. It is planned to be used as armour tiles for the divertor or the first wall. It turned out that no specific effort has been undertaken so far to search for a suitable irradiation location in the IFMIF Test Cell which provides a reasonable representation of the irradiation conditions in the divertor of a fusion power reactors. (orig.)

Comparison of properties and microstructures of Trefimetaux CuNiBe and Hycon 3HP {sup TM} before and after neutron irradiation. (ITER R and D Task no. T213)

Energy Technology Data Exchange (ETDEWEB)

Edwards, D.J. [Pacific Northwest National Lab., Materials Development. Group, Richland (United States); Eldrup, M.; Toft, P.; Singh, B.N

2000-07-01

The precipitation strengthened CuNiBe alloys are among the three candidate copper alloys that are 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 Cu-A1{sub 2}O{sub 3} alloys. Brush-Wellman Inc. has developed an improved version of their Hycon CuNiBe alloy that has higher conductivity while maintaining a reasonable level of strength. In the present work we have investigated the physical and mechanical properties of the Hycon 3HP{sup TM} alloy both before and after neutron irradiation and have compared its microstructure and properties with the European CuNiBe candidate alloy manufactured by Trefimetaux. Tensile specimens of both alloys were irradiated in the DR-3 reactor at Risoe to displacement dose levels of up to 0.3 dpa at 100, 250 and 350 {sup d}eg{sup C}. Both alloys were tensile tested in vacuum in the unirradiated and irradiated conditions at 100, 250 and 350 {sup d}eg{sup C} and the microstructures of the alloys were investigated using transmission electron microscopy. Electrical resistivity measurements were made on tensile specimens be-fore and after irradiation; all measurements were made at 23 {sup d}eg{sup C}. Results of these investigations are presented and discussed in terms of the sensitivity of these alloys to test temperature, which becomes increasingly problematic when the irradiation and test temperature reaches 250 {sup d}eg{sup C} and above. (au)

EXPERIMENTAL INVESTIGATION ON TENSILE STRENGTH OF JACQUARD KNITTED FABRICS

Directory of Open Access Journals (Sweden)

BRAD Raluca

2015-05-01

Full Text Available An objective approach to select the best fabric for technical and home textiles consists in mechanical properties evaluation. The goal of this study is to analyze the behavior of knitted fabrics undergoing stretch stress. In this respect, three types of 2 colors Rib structure (backstripes jacquard, twillback jacquard and double-layered 3x3 rib fabric have been presented and tested for tensile strength and elongation on three directions. First, the elasticity and the behavior of knitted Rib fabrics were described The fabrics were knitted using 100% PAN yarns with Nm 1/15x2 on a E5 CMS 330 Stoll V-bed knitting machine, and have been tested using INSTROM 5587 Tensile Testing Machine in respect of standards conditions. After a relaxation period, 15 specimens were prepared, being disposed at 0°, 45 and 90 angles to the wale direction on the flat knitted panel. The tensile strength and the elongation values were recorded and mean values were computed. After strength and tensile elongation testing for 3 types of rib based knitted fabrics, one can see that the double layer knit presents the best mechanical behavior, followed by birds-eyebacking 2 colors Jacquard and then back striped Jacquard. For tensile stress in bias direction, the twillbacking Jacquard has a good breakage resistance value due to the higher number of rib sinker loops in structure that are positioned on the same direction with the tensile force. The twillbacking Jacquard structure could be considered as an alternative for the base material for decorative and home textile products.

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI Series 5

International Nuclear Information System (INIS)

Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O.; Menke, B.H.

1992-10-01

The Fifth Irradiation Series in the Heavy-Section Steel Irradiation Program obtained a statistically significant fracture toughness data base on two high-copper (0.23 and 0.31 wt %) submerged-arc welds to determine the shift and shape of the K Ic curve as a consequence of irradiation. Compact specimens with thicknesses to 101.6 mm (4 in) in the irradiated condition and 203.2 mm (8 in) in the unirradiated condition were tested, in addition to Charpy impact, tensile, and drop-weight specimens. Irradiations were conducted at a nominal temperature of 288 degree C and an average fluence of 1.5 x 10 19 neutrons/cm 2 (>l MeV). The Charpy 41-J temperature shifts are about the same as the corresponding drop-weight NDT temperature shifts. The irradiated welds exhibited substantial numbers of cleavage pop-ins. Mean curve fits using two-parameter (with fixed intercept) nonlinear and linearized exponential regression analysis revealed that the fracture toughness 100 MPa lg-bullet √m shifts exceeded the Charpy 41-J shifts for both welds. Analyses of curve shape changes indicated decreases in the slopes of the fracture toughness curves, especially for the higher copper weld. Weibull analyses were performed to investigate development of lower bound curves to the data, including the use of a variable K min parameter which affects the curve shape

Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

Energy Technology Data Exchange (ETDEWEB)

Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

2011-07-01

The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J.; Alexander, D.J.; Robertson, J.P. [Oak Ridge National Lab., TN (United States)] [and others

1997-04-01

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

Patients exposed to therapeutic irradiation for benign conditions in childhood

International Nuclear Information System (INIS)

Carroll, R.G.

1976-01-01

In the United States there are over 1 million individuals who have been irradiated for benign conditions in the head and neck. Many recent reports document conclusively that individuals who have had head and neck radiation for benign conditions have markedly increased risk of thyroid, salivary, and perhaps breast cancer as compared to the general population. Although the relative risk as compared to the general population is high, the risk that any one individual who has had head or neck irradiation will develop a subsequent malignancy is relatively low. Identification of these patients through some type of screening procedure may be beneficial in terms of prevention of subsequent mortality and morbidity from cancer, especially thyroid and salivary cancer. The American Medical Association and the American Hospital Association issued a joint statement of October 17, 1975, urging hospitals and physicians to work together in their communities to develop guidelines and procedures for screening of exposed individuals and for public education

Available post-irradiation examination techniques at Romanian institute for nuclear research

International Nuclear Information System (INIS)

Parvan, Marcel; Sorescu, Antonius; Mincu, Marin; Uta, Octavian; Dobrin, Relu

2005-01-01

The Romanian Institute for Nuclear Research (INR) has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The available techniques of post-irradiation examination (PIE) and purposes of PIE for CANDU reactor fuel are as follows. 1) Visual inspection and photography by periscope: To examine the surface condition such as deposits, corrosion etc. 2) Eddy current testing: To verify the cladding integrity. 3) Profilometry and length measurement performed both before and after irradiation: To measure the parameters which highlight the dimensional changes i.e. diameter, length, diametral and axial sheath deformation, circumferential sheath ridging height, bow and ovality. 4) Gamma scanning and Tomography: To determine the burnup, axial and radial fission products activity distribution and to check for flux peaking and loading homogeneity. 5) Puncture test: To measure the pressure, volume and composition of fission gas and the inner free volume. 6) Optical microscopy: To highlight the structural changes and hydriding, to examine the condition of the fuel-sheath interface and to measure the oxide thickness and Vickers microhardness. 7) Mass spectrometry: To measure the burnup. 8) Tensile testing: To check the mechanical properties. So far, non-destructive and destructive post-irradiation examinations have been performed on a significant number of CANDU fuel rods (about 100) manufactured by INR and irradiated to different power histories in the INR 14 MW(th) TRIGA reactor. These examinations have been performed as part of the Romanian research programme for the manufacturing, development and safety of the CANDU fuel. The paper describes the PIE techniques and some results. (Author)

Computational science simulation of laser materials processing and provision of their irradiation conditions

International Nuclear Information System (INIS)

Muramatsu, Toshiharu

2016-01-01

In laser processing, it is necessary for achieving the intended performance and product, to understand the complex physical courses including melting and solidification phenomena occurring in laser processing, and thus to set proper laser irradiation conditions. This condition optimization work requires an enormous amount of overhead due to repeated efforts, and has become a cause for inhibiting the introduction of laser processing technology into the industrial field that points to the small lot production of many products. JAEA tried to make it possible to quantitatively handle the complex physical course from the laser light irradiation to the fabricating material until the completion of processing, and is under development of the computational science simulation code SPLICE that connects micro behavior and macro behavior through a multi-level scale model. This SPLICE is able to visualize the design space and to reduce the overhead associated with the setting of laser irradiation conditions and the like, which gives the prospect of being effective as a tool for front-loading. This approach has been confirmed to be effective for the welding and fusing process. (A.O.)

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

Irradiation combined with Bleomycin treatment of synchronized cells in culture under oxic and hypoxic conditions

International Nuclear Information System (INIS)

Midander, J.; Littbrand, B.; Edsmyr, F.

1980-01-01

Bleomycin-treated cells are sensitized to radiation delivered under oxic conditions both in the early S and G 2 phases of the cycle, irradiated under hypoxic conditions, sensitization occurs only in the early S phase. This difference in the sensitizing effect of the drug is discussed in regard to the possible clinical advantages of a combined treatment of tumours with irradiation and Bleomycin. (Auth.)

Mechanical Properties and Microstructure of Neutron Irradiated Cold-worked Al-1050 and Al-6063 Alloys

International Nuclear Information System (INIS)

Munitz, A.; Cotler, A; Talianker, M.

1998-01-01

The impact of neutron irradiation on the internal microstructure, mechanical properties and fracture morphology of cold-worked Al-1050 and Al-6063 alloys was studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens consisting of 50 mm long and 6 mm wide gauge sections, were punched out from Al-1050 and Al-6063 23% cold-worked tubes. They were exposed to prolonged neutron irradiation of up to 4.5x10 25 and 8x10 25 thermal neutrons/m 2 (E -3 s -1 . In general, the uniform and total elongation, the yield stress, and the ultimate tensile strength increase as functions of fluence. However, for Al-1050 a decrease in the ultimate tensile strength and yield stress was observed up to a fluence of 1x10 25 thermal neutrons/m 2 which then increase with thermal neutrons fluence. Metallographic examination and fractography for Al-6063 revealed a decrease in the local area reduction of the final fracture necking. This reduction is accompanied by a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. In contrast, for Al-1050, fracture morphology remains ductile transgranular shear rupture and the final local area reduction remains almost constant No voids could be observed in either alloy up to the maximum fluence. The dislocation density of cold-worked Al was found to decrease with the thermal neutron fluence. Prolonged annealing of unirradiated cold-worked Al-6063 at 52 degree led to similar results. Thus, it appears that, under our irradiation conditions, whereby the temperature encompassing the samples increases the exposure to this thermal field is the major factor influencing the mechanical properties and microstructure of aluminum alloys

Embrittlement of a 17Cr ferritic steel irradiated in Phenix

International Nuclear Information System (INIS)

Allegraud, G.; Boutard, J.L.; Boyer, J.M.

1987-01-01

Charpy V and tensile tests have been performed with samples made of 17Cr ferritic steel irradiated in Phenix at temperatures between 390 and 540C up to a maximum dose of 83.3 dpaF. All over the temperature and dose ranges, irradiation leads to an increase of the ductile brittle transition temperature (DBTT). The DBTT and hardening are decreasing functions of the irradiation temperature. Fast neutron flux at 390C hardens the material more than a sole thermal ageing does

Poly (Lactic Acid)/Layered Silicate Nanocomposite Films: Effect of Irradiation

Energy Technology Data Exchange (ETDEWEB)

Dadbin, S.; Naimian, F.; Akhavan, A.; Hasanpoor, S., E-mail: sdadbin@yahoo.com, E-mail: sdadbin@aeoi.org.ir [Atomic Energy Organization of Iran (AEOI), Nuclear Science and Technology Research Institute, P.O. Box 11365-8486, Tehran, North Kargar (Iran, Islamic Republic of)

2010-07-01

Poly (Lactic acid) –layered silicate nanocomposite films were prepared by solution casting method. The films were irradiated with Co{sup 60} radiation facility at dose of 30 kGy. The effect of gamma irradiation on mechanical properties of the neat PLA and nanocomposites was evaluated by data obtained from tensile testing measurements. The tensile strength of the irradiated PLA films increased with addition of 1 wt% Triallyl Cyanurate (TAC) indicating crosslink formation. Significant ductile behavior was observed in the PLA nanocomposites containing 4 pph of nanoclay. Incorporation of nanoclay particles in the PLA matrix stimulated crystal growth as it was studied by differential scanning calorimetry (DSC). The morphology of the nanocomposites characterized by transmission electron microscopy (TEM) and X- ray diffraction (XRD) revealed an exfoliated morphology in the PLA nanocomposite films containing 4 pph of nanoclay. Only very small changes were observed in the chemical structure of the irradiated samples as it was investigated by Fourier transform infrared (FTIR) spectroscopy. Enzymatic degradation rate of PLA and its nanocomposite decreased with increasing crystallinity of the samples. The rate of weight loss was also affected by the morphology of the nanocomposites. (author)

Poly (Lactic Acid)/Layered Silicate Nanocomposite Films: Effect of Irradiation

International Nuclear Information System (INIS)

Dadbin, S.; Naimian, F.; Akhavan, A.; Hasanpoor, S.

2010-01-01

Poly (Lactic acid) –layered silicate nanocomposite films were prepared by solution casting method. The films were irradiated with Co 60 radiation facility at dose of 30 kGy. The effect of gamma irradiation on mechanical properties of the neat PLA and nanocomposites was evaluated by data obtained from tensile testing measurements. The tensile strength of the irradiated PLA films increased with addition of 1 wt% Triallyl Cyanurate (TAC) indicating crosslink formation. Significant ductile behavior was observed in the PLA nanocomposites containing 4 pph of nanoclay. Incorporation of nanoclay particles in the PLA matrix stimulated crystal growth as it was studied by differential scanning calorimetry (DSC). The morphology of the nanocomposites characterized by transmission electron microscopy (TEM) and X- ray diffraction (XRD) revealed an exfoliated morphology in the PLA nanocomposite films containing 4 pph of nanoclay. Only very small changes were observed in the chemical structure of the irradiated samples as it was investigated by Fourier transform infrared (FTIR) spectroscopy. Enzymatic degradation rate of PLA and its nanocomposite decreased with increasing crystallinity of the samples. The rate of weight loss was also affected by the morphology of the nanocomposites. (author)

Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

Energy Technology Data Exchange (ETDEWEB)

Rabenberg, Ellen M.; Jaques, Brian J. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Sencer, Bulent H. [Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garner, Frank A. [Radiation Effects Consulting, 2003 Howell Ave., Richland, WA 99354 (United States); Freyer, Paula D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Okita, Taira [Research Into Artifacts Dept., Center for Engineering, University of Tokyo, Tokyo (Japan); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

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.

Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

International Nuclear Information System (INIS)

Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

1981-01-01

Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700 0 C to fluence of 3 x 10 26 n/m 2 , E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700 0 C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations

Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

Energy Technology Data Exchange (ETDEWEB)

Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

1981-01-01

Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700/sup 0/C to fluence of 3 x 10/sup 26/ n/m/sup 2/, E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700/sup 0/C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations.

Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

International Nuclear Information System (INIS)

Garlick, A.; Hindmarch, P.

1980-09-01

Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 deg. C

International Nuclear Information System (INIS)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-01-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 deg. C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 deg. C and 300 deg. C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 deg. C (up to 2.6 dpa), and tested between -170 deg. C and 300 deg. C. Irradiation effects at lower irradiation temperatures are more significant

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

Science.gov (United States)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-06-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

The effects of combined treatment of irradiation and heat on bacteria escherichia coli and sarcina lutea in dry condition

International Nuclear Information System (INIS)

Nikham; Hilmy, Nazly

1987-01-01

The effects of combination treatment of irradiation and heat on bacteria escherichia coli and sarcina lutea in dry condition. Investigation on the effects of combined irradiation + heat and heat + irradiation treatments have been carried out i.e. at the doses of 0; 1.0; 1.5; and 2.0 kGy with heating at 50 0 C for 10; 20; and 30 minutes on escherichia coli B/r, escherichia coli from sludge and sarcine lutea. Samples of bacteria were prepared in dry condition by using sterile fine sand as carrier. Irradiation was done in aerobic condition with RH 90% and the time range between irradiation and heating was not more than 2 hours. The results showed that the D 10 value did not give significant difference between the combined irradiation + heat, and heat + irradiation treatments for the 3 species of bacteria, compared to irradiation only (p 0.05). Doses of 1.0 and 1.5 kGy combined with heating at 50 0 C for 10 and 20 minutes gave better results compared to irradiation only. 17 refs

Mechanical properties of irradiated 9Cr-2WVTa steel

International Nuclear Information System (INIS)

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

1998-01-01

An Fe-9Cr-2W-0.25V-0.07Ta-0.1C (9Cr-2WVTa) steel has excellent strength and impact toughness before and after irradiation in the Fast Flux Test Facility and the High Flux Reactor (HFR). The ductile-brittle transition temperature (DBTT) increased only 32 C after 28 dpa at 365 C in FFTF, compared to a shift of ∼60 C for a 9Cr-2WV steel--the same as the 9Cr-2WVTa steel but without tantalum. This difference occurred despite the two steels having similar tensile but without tantalum. This difference occurred despite the two steels having similar tensile properties before and after irradiation. The 9Cr-2WVTa steel has a smaller prior-austenite grain size, but otherwise microstructures are similar before irradiation and show similar changes during irradiation. The irradiation behavior of the 9Cr-2WVTa steel differs from the 9Cr-2WV steel and other similar steels in two ways: (1) the shift in DBTT of the 9Cr-2WVTa steel irradiated in FFTF does not saturate with fluence by ∼28 dpa, whereas for the 9Cr-2WV steel and most similar steels, saturation occurs at <10 dpa, and (2) the shift in DBTT for 9Cr-2WVTa steel irradiated in FFTF and HFR increased with irradiation temperature, whereas it decreased for the 9Cr-2WV steel, as it does for most similar steels. The improved properties of the 9Cr-2WVTa steel and the differences with other steels were attributed to tantalum in solution

Development and characterization of biodegradable polymer blends - PHBV/PCL irradiated with gamma rays

International Nuclear Information System (INIS)

Rosario, F.; Casarin, S.A.; Agnelli, J.A.M.; Souza Junior, O.F. de

2010-01-01

This paper presents the results of a study that aimed to develop PHBV biodegradable polymer blends, in a major concentration with PCL, irradiate the pure polymers and blends in two doses of gamma radiation and to analyze the changes in chemical and mechanical properties. The blends used in this study were from natural biodegradable copolymer poly (hydroxybutyrate-valerate) (PHBV) and synthetic biodegradable polymer poly (caprolactone) (PCL 2201) with low molar mass (2,000 g/mol). Several samples were prepared in a co-rotating twin-screw extruder and afterwards, the tensile specimens were injected for the irradiation treatment with 50 kGy to 100 kGy doses and for the mechanical tests. The characterization of the samples before and after the irradiation treatments was performed through scanning electron microscopy (SEM), dynamic mechanical thermal analysis (DMTA), differential scanning calorimetry (DSC) and mechanical tensile tests. (author)

Changes in quality of apple (Malus domestica) cultivars due to γ-irradiation and storage conditions

International Nuclear Information System (INIS)

Hussain, P.R.; Dar, M.A.; Meena, R.S.; Wani, A.M.; Mir, M.A.; Shafi, F.

2008-01-01

Ambri, Golden Delicious and Royal Delicious apple varieties were γ-irradiated (0.1-0.5 kGy) and stored under ambient (15 ±2 degC, 80%RH) and refrigerated (3±1 degC, 90%RH) conditions. They were evaluated periodically for firmness, total soluble solids, acidity, juice yield, physiological loss in weight, overall acceptability and yeast and mold counts. The γ-irradiation doses of 0.2, 0.3 and 0.5 kGy proved beneficial in maintaining the overall quality of all the 3 varieties of apple under both the storage conditions. γ-irradiation significantly reduced the yeast and mold counts of apples under storage. (author)

Irradiation behavior of German PWR RPV steels under operating conditions

Energy Technology Data Exchange (ETDEWEB)

May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

2011-07-01

In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

Preparation of pinewood/polymer/composites using gamma irradiation

Energy Technology Data Exchange (ETDEWEB)

Ajji, Zaki [Polymer Technology Division, Department of Radiation Technology, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)]. E-mail: atomic@aec.org.sy

2006-09-15

Wood/polymer composites (WPC) have been prepared from pinewood with different compounds using gamma irradiation: butyl acrylate, butyl methacrylate, styrene, acrylamide, acrylonitrile, and unsaturated polyester styrene resin. The polymer loading was determined with respect to the compound concentration and the irradiation dose. The polymer loading increases generally with increase in the monomer or polymer concentration. Tensile and compression strength have been improved in the four cases, but no improvement was observed using unsaturated polyester styrene resin or acrylamide.

Physical properties of beryllium oxide - Irradiation effects

International Nuclear Information System (INIS)

Elston, J.; Caillat, R.

1958-01-01

This work has been carried out in view of determining several physical properties of hot-pressed beryllium oxide under various conditions and the change of these properties after irradiation. Special attention has been paid on to the measurement of the thermal conductivity coefficient and thermal diffusivity coefficient. Several designs for the measurement of the thermal conductivity coefficient have been achieved. They permit its determination between 50 and 300 deg. C, between 400 and 800 deg. C. Some measurements have been made above 1000 deg. C. In order to measure the thermal diffusivity coefficient, we heat a perfectly flat surface of a sample in such a way that the heat flux is modulated (amplitude and frequency being adjustable). The thermal diffusivity coefficient is deduced from the variations of temperature observed on several spots. Tensile strength; compressive strength; expansion coefficient; sound velocity and crystal parameters have been also measured. Some of the measurements have been carried out after neutron irradiation. Some data have been obtained on the change of the properties of beryllium oxide depending on the integrated neutron flux. (author) [fr

2D micromechanical analysis of SiC/Al metal matrix composites under tensile, shear and combined tensile/shear loads

DEFF Research Database (Denmark)

Qing, Hai

2013-01-01

The influence of interface strength and loading conditions on the mechanical behavior of the metal-matrix composites is investigated in this paper. A program is developed to generate automatically 2D micromechanical Finite element (FE) models including interface, in which both the locations...... and dimensions of Silicon-Carbide (SiC) particles are randomly distributed. Finite element simulations of the deformation and damage evolution of SiC particle reinforced Aluminum (Al) alloy composite are carried out for different microstructures and interphase strengths under tensile, shear and combined tensile....../shear loads. 2D cohesive element is applied to describe the fracture and failure process of interphase, while the damage models based on maximum principal stress criterion and the stress triaxial indicator are developed within Abaqus/Standard Subroutine USDFLD to simulate the failure process of SiC particles...

Characterization of gamma irradiated peanut kernels stored one year under ambient and frozen conditions

International Nuclear Information System (INIS)

Chiou, R.Y.Y.; Shyu, S.L.; Tsai, C.L.

1991-01-01

Peanut kernels were gamma irradiated at 0, 2.5, 5.0, 10, and 20 KGy, and stored 1 yr at ambient and frozen (-14 degrees C) conditions. Irradiated peanuts lost germination capabilities during storage. Molds were detected only on peanuts irradiated with 2.5 KGy and stored at ambient temperature. Peanut oil in kernels stored at -14 degrees C was comparatively more stable than that in peanuts stored at ambient temperature. Oxidation of oil was not significantly changed by irradiation. Changes in fatty acid content varied slightly with exception of linoleic and linolenic acids which decreased with increased radiation depending on storage temperature. The SDS-PAGE protein patterns of peanuts revealed no noticeable variation of protein subunits resulting from irradiation and storage

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

Loelgen, R.H.; Cundy, M.R.; Schuele, W.

1977-01-01

A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

Science.gov (United States)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-03-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 × 1026 n/m2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 × 1026 n/m2. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

International Nuclear Information System (INIS)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-01-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

Effect of irradiation temperature and strain rate on the mechanical properties of V-4Cr-4Ti irradiated to low doses in fission reactors

International Nuclear Information System (INIS)

Zinkle, S.J.; Snead, L.L.; Rowcliffe, A.F.; Alexander, D.J.; Gibson, L.T.

1998-01-01

Tensile tests performed on irradiated V-(3-6%)Cr-(3-6%)Ti alloys indicate that pronounced hardening and loss of strain hardening capacity occurs for doses of 0.1--20 dpa at irradiation temperatures below ∼330 C. The amount of radiation hardening decreases rapidly for irradiation temperatures above 400 C, with a concomitant increase in strain hardening capacity. Low-dose (0.1--0.5 dpa) irradiation shifts the dynamic strain aging regime to higher temperatures and lower strain rates compared to unirradiated specimens. Very low fracture toughness values were observed in miniature disk compact specimens irradiated at 200--320 C to ∼1.5--15 dpa and tested at 200 C

Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

International Nuclear Information System (INIS)

Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

2011-01-01

Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

Conditioned instrumental behaviour in the rat: Effects of prenatal irradiation with various low dose-rate doses

International Nuclear Information System (INIS)

Klug, H.

1986-01-01

4 groups of rats of the Wistar-strain were subjected to γ-irradiation on the 16th day of gestation. 5 rats received 0,6 Gy low dose rate irradiation, 5 animals received 0,9 Gy low dose and 6 high dose irradiation, 3 females were shamirradiated. The male offspring of these 3 irradiation groups and 1 control group were tested for locomotor coordination on parallel bars and in a water maze. The female offspring were used in an operant conditioning test. The locomotor test showed slight impairment of locomotor coordination in those animals irradiated with 0,9 Gy high dose rate. Swimming ability was significantly impaired by irradiation with 0,9 Gy high dose rate. Performance in the operant conditioning task was improved by irradiation with 0,9 Gy both low and high dose rate. The 0,9 Gy high dose rate group learned faster than all the other groups. For the dose of 0,9 Gy a significant dose rate effect could be observed. For the dose of 0,6 Gy a similar tendency was observed, differences between 0,6 Gy high and low dose rate and controls not being significant. (orig./MG) [de

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

International Nuclear Information System (INIS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H.W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-01-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2x10 22 m -2 (E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

Science.gov (United States)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-12-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

International Nuclear Information System (INIS)

Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

2008-01-01

An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

The effects of fast-neutron irradiation on the mechanical properties of austenitic stainless steel

International Nuclear Information System (INIS)

Dalton, J.H.

1978-01-01

The paper reviews the effects of fast-neutron irradiation on the tensile properties of austenitic stainless steels at irradiation temperatures of less than 400 degrees Celcius, using as an example, work carried out at Pelindaba on an AISI 316 type steel produced in South Africa. Damage produced in these steels at higher irradiation temperatures and fluences is also briefly discussed. The paper concludes with a discussion of some methods of overcoming or decreasing the effects of irradiation damage [af

Mechanical properties of irradiated materials

International Nuclear Information System (INIS)

Robertson, I.M.; Robach, J.; Wirth, B.

2001-01-01

The effect of irradiation on the mechanical properties of metals is considered with particular attention being paid to the development of defect-free channels following uniaxial tensile loading. The in situ transmission electron microscope deformation technique is coupled with dislocation dynamic computer simulations to reveal the fundamental processes governing the elimination of defects by glissile dislocations. The observations of preliminary experiments are reported.(author)

Investigation of microstructure and mechanical properties of proton irradiated Zircaloy 2

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Apu, E-mail: asarkar@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Kumar, Ajay [Nuclear Physics Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Mukherjee, S.; Sharma, S.K.; Dutta, D.; Pujari, P.K. [Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Agarwal, A.; Gupta, S.K.; Singh, P. [Ion Accelerator Development Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India)

2016-10-15

Samples of Zircaloy 2 have been irradiated with 4 MeV protons to two different doses. Microstructures of the unirradiated and irradiated samples have been characterized by Electron Back Scatter Diffraction (EBSD), X-ray diffraction line profile analysis (XRDLPA), Positron Annihilation Lifetime Spectroscopy (PALS) and Coincident Doppler Broadening (CDB) Spectroscopy. Tensile tests and micro hardness measurements have been carried out at room temperature to assess the changes in mechanical properties of Zircaloy 2 due to proton irradiation. The correlation of dislocation density, grain size and yield stress of the irradiated samples indicated that an increase in dislocation density due to irradiation is responsible for the change in mechanical behavior of irradiated Zircaloy.

Laser-induced generation of pure tensile stresses

International Nuclear Information System (INIS)

Niemz, M.H.; Lin, C.P.; Pitsillides, C.; Cui, J.; Doukas, A.G.; Deutsch, T.F.

1997-01-01

While short compressive stresses can readily be produced by laser ablation, the generation of pure tensile stresses is more difficult. We demonstrate that a 90 degree prism made of polyethylene can serve to produce short and pure tensile stresses. A compressive wave is generated by ablating a thin layer of strongly absorbing ink on one surface of the prism with a Q-switched frequency-doubled Nd:YAG laser. The compressive wave driven into the prism is reflected as a tensile wave by the polyethylene-air interface at its long surface. The low acoustic impedance of polyethylene makes it ideal for coupling tensile stresses into liquids. In water, tensile stresses up to -200bars with a rise time of the order of 20 ns and a duration of 100 ns are achieved. The tensile strength of water is determined for pure tensile stresses lasting for 100 ns only. The technique has potential application in studying the initiation of cavitation in liquids and in comparing the effect of compressive and tensile stress transients on biological media. copyright 1997 American Institute of Physics

Effect of gamma irradiation on physicochemical properties of commercial poly(lactic acid) clamshell for food packaging

International Nuclear Information System (INIS)

Madera-Santana, Tomás J.; Meléndrez, R.; González-García, Gerardo; Quintana-Owen, Patricia; Pillai, Suresh D.

2016-01-01

Poly(lactic acid) (PLA) is a well-known biodegradable polymer with strong potential application in food packaging industry. In this paper, samples of PLA clamshell for tomatoes packaging were exposed with 60 CO γ-ray's source (1.33 MeV) at different dose levels (0, 10, 60, 150, 300, and 600 kGy), at room temperature and in presence of air. The physicochemical properties of neat PLA and sample exposed to gamma irradiation were investigated using Fourier transform infrared spectroscopy (FTIR), X-ray diffraction (XRD), gel permeation chromatography (GPC), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), scanning electron microscopy (SEM) and tensile measurements. Results show as the dose increases, the molecular weight (M w ), melting temperature (T m ), tensile strength and elongation at break decreased. However, the tensile modulus increased with increasing doses. The surface of PLA clamshells was degraded (scratches and minor cracks) when samples were exposed to doses greater than 60 kGy. - Highlights: • The gamma irradiation effects on PLA clamshells were studied. • DSC, XRD, NMR and FTIR analysis were used for PLA clamshell characterization. • The M w , T m , strength and elongation of the irradiated PLA clamshells decreased. • The tensile modulus increased with increasing gamma doses. • The Surface of PLA clamshell showed scratches and minor cracks.

Laser irradiation effects on the surface, structural and mechanical properties of Al-Cu alloy 2024

Science.gov (United States)

Yousaf, Daniel; Bashir, Shazia; Akram, Mahreen; kalsoom, Umm-i.-; Ali, Nisar

2014-02-01

Laser irradiation effects on surface, structural and mechanical properties of Al-Cu-Mg alloy (Al-Cu alloy 2024) have been investigated. The specimens were irradiated for various fluences ranging from 3.8 to 5.5 J/cm2 using an Excimer (KrF) laser (248 nm, 18 ns, 30 Hz) under vacuum environment. The surface and structural modifications of the irradiated targets have been investigated by scanning electron microscope (SEM) and X-ray diffractometer (XRD), respectively. SEM analysis reveals the formation of micro-sized craters along the growth of periodic surface structures (ripples) at their peripheries. The size of the craters initially increases and then decreases by increasing the laser fluence. XRD analysis shows an anomalous trend in the peak intensity and crystallite size of the specimen irradiated for various fluences. A universal tensile testing machine and Vickers microhardness tester were employed in order to investigate the mechanical properties of the irradiated targets. The changes in yield strength, ultimate tensile strength and microhardness were found to be anomalous with increasing laser fluences. The changes in the surface and structural properties of Al-Cu alloy 2024 after laser irradiation have been associated with the changes in mechanical properties.

Experimental Investigation for Tensile Performance of GFRP-Steel Hybridized Rebar

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Dong-Woo Seo

2016-01-01

Full Text Available Tensile performance of the recently developed “FRP Hybrid Bar” at Korea Institute of Civil Engineering and Building Technology (KICT is experimentally evaluated by the authors. FRP Hybrid Bar is introduced to overcome the low elastic modulus of the existing GFRP bars to be used as a structural member in reinforced concrete structures. The concept of material hybridization is applied to increase elastic modulus of GFRP bars by using steel. This hybridized GFRP bar can be used in concrete structures as a flexural reinforcement with a sufficient level of elastic modulus. In order to verify the effect of material hybridization on tensile properties, tensile tests are conducted. The test results for both FRP Hybrid Bar and the existing GFRP bars are compared. The results indicate that the elastic modulus of FRP Hybrid Bar can be enhanced by up to approximately 250 percent by the material hybridization with a sufficient tensile strength. To ensure the long-term durability of FRP Hybrid Bar to corrosion resistance, the individual and combined effects of environmental conditions on FRP Hybrid Bar itself as well as on the interface between rebar and concrete are currently under investigation.

The effect of different light-curing units on tensile strength and microhardness of a composite resin

Directory of Open Access Journals (Sweden)

Eduardo Batista Franco

2007-12-01

Full Text Available The aim of this study was to evaluate the influence of different light-curing units on the tensile bond strength and microhardness of a composite resin (Filtek Z250 - 3M/ESPE. Conventional halogen (Curing Light 2500 - 3M/ESPE; CL and two blue light emitting diode curing units (Ultraled - Dabi/Atlante; UL; Ultrablue IS - DMC; UB3 and UB6 were selected for this study. Different light intensities (670, 130, 300, and 600 mW/cm², respectively and different curing times (20s, 40s and 60s were evaluated. Knoop microhardness test was performed in the area corresponding to the fractured region of the specimen. A total of 12 groups (n=10 were established and the specimens were prepared using a stainless steel mold composed by two similar parts that contained a cone-shaped hole with two diameters (8.0 mm and 5.0 mm and thickness of 1.0 mm. Next, the specimens were loaded in tensile strength until fracture in a universal testing machine at a crosshead speed of 0.5 mm/min and a 50 kg load cell. For the microhardness test, the same matrix was used to fabricate the specimens (12 groups; n=5. Microhardness was determined on the surfaces that were not exposed to the light source, using a Shimadzu HMV-2 Microhardness Tester at a static load of 50 g for 30 seconds. Data were analyzed statistically by two-way ANOVA and Tukey's test (p<0.05. Regarding the individual performance of the light-curing units, there was similarity in tensile strength with 20-s and 40-s exposure times and higher tensile strength when a 60-s light-activation time was used. Regarding microhardness, the halogen lamp had higher results when compared to the LED units. For all light-curing units, the variation of light-exposure time did not affect composite microhardness. However, lower irradiances needed longer light-activation times to produce similar effect as that obtained with high-irradiance light-curing sources.

Dynamic tensile behaviour and deformational mechanism of C5191 phosphor bronze under high strain rates deformation

Energy Technology Data Exchange (ETDEWEB)

Hu, Dao-chun [College of Mechanical and Electrical Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); College of Mechanical and Electrical Engineering, Taizhou Vocational & Technical College, Taizhou 318000 (China); Chen, Ming-he, E-mail: meemhchen@nuaa.edu.cn [College of Mechanical and Electrical Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Wang, Lei; Cheng, Hu [College of Mechanical Engineering, Taizhou University, Taizhou 318000 (China)

2016-01-01

High speed stamping process is used to high strength and high electrical conductivity phosphor bronze with extremely high strain rates more than 10{sup 3} s{sup −1}. This study on the dynamic tensile behaviour and deformational mechanism is to optimise the high speed stamping processes and improve geometrical precision in finished products. Thus, the tensile properties and deformation behaviour of C5191 phosphor bronze under quasi-static tensile condition at a strain rate of 0.001 s{sup −1} by electronic universal testing machine, and dynamic tensile condition at strain rate of 500, 1000 and 1500 s{sup −1} by split Hopkinson tensile bar (SHTB) apparatus were studied. The effects of strain rate and the deformation mechanism were investigated by means of SEM and TEM. The results showed that the yield strength and tensile strength of C5191 phosphor bronze under high strain rates deformation increased by 32.77% and 11.07% respectively compared with quasi-static condition, the strain hardening index increases from 0.075 to 0.251, and the strength of the material strain rates sensitivity index change from 0.005 to 0.022, which presented a clear sensitive to strain rates. Therefore, it is claimed that the dominant deformation mechanism was changed by the dislocation motion under different strain rates, and the ability of plastic deformation of C5191 phosphor bronze increased due to the number of movable dislocations increased significantly, started multi-line slip, and the soft effect of adiabatic temperature rise at the strain rate ranging from 500 to 1500 s{sup −1}.

Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

2008-08-01

The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

Trialed production of low protein irradiated natural rubber latex in factory scale by gamma irradiation technique

International Nuclear Information System (INIS)

Utama, Marga; Herwinarni, S.; Halik, H.M.; Siswanto; Suharyanto; Syamsu, Y.; Handoko, B.

2006-01-01

Four tons fresh field natural rubber latex (FNRL) with total solid content 30% were added with 2 phr (part hundred ratio of rubber) normal butyl acrylate (nBA) then irradiated by gamma rays at 25 kGy. The irradiated FNRL was centrifuged, then the properties of irradiated centrifuged natural rubber latex (INRL) and its film were measured before and after storage for 5 months. It is found that the INRL is stable latex during storage in 5 months, with lowest protein, and free nitrosamine content. The tensile strength of INRL film was 24-27 MPa, and modulus 600% was 1,5-2,0 MPa, elongation at break was 900%, and hardness was 27-29 Shore A, while the extractable protein content less than 100 μg/g. (author)

The effect of gamma-irradiation conditions on the immunogenicity of whole-inactivated Influenza A virus vaccine.

Science.gov (United States)

David, Shannon C; Lau, Josyane; Singleton, Eve V; Babb, Rachelle; Davies, Justin; Hirst, Timothy R; McColl, Shaun R; Paton, James C; Alsharifi, Mohammed

2017-02-15

Gamma-irradiation, particularly an irradiation dose of 50kGy, has been utilised widely to sterilise highly pathogenic agents such as Ebola, Marburg Virus, and Avian Influenza H5N1. We have reported previously that intranasal vaccination with a gamma-irradiated Influenza A virus vaccine (γ-Flu) results in cross-protective immunity. Considering the possible inclusion of highly pathogenic Influenza strains in future clinical development of γ-Flu, an irradiation dose of 50kGy may be used to enhance vaccine safety beyond the internationally accepted Sterility Assurance Level (SAL). Thus, we investigated the effect of irradiation conditions, including high irradiation doses, on the immunogenicity of γ-Flu. Our data confirm that irradiation at low temperatures (using dry-ice) is associated with reduced damage to viral structure compared with irradiation at room temperature. In addition, a single intranasal vaccination with γ-Flu irradiated on dry-ice with either 25 or 50kGy induced seroconversion and provided complete protection against lethal Influenza A challenge. Considering that low temperature is expected to reduce the protein damage associated with exposure to high irradiation doses, we titrated the vaccine dose to verify the efficacy of 50kGy γ-Flu. Our data demonstrate that exposure to 50kGy on dry-ice is associated with limited effect on vaccine immunogenicity, apparent only when using very low vaccine doses. Overall, our data highlight the immunogenicity of influenza virus irradiated at 50kGy for induction of high titre antibody and cytotoxic T-cell responses. This suggests these conditions are suitable for development of γ-Flu vaccines based on highly pathogenic Influenza A viruses. Copyright © 2017 Elsevier Ltd. All rights reserved.

Effects of pre-irradiation conditioning of Medfly pupae (Diptera: Tephritidae): Hypoxia and quality of sterile males

International Nuclear Information System (INIS)

Nestel, D.; Nemny-Lavy, E.; Islam, S.M.; Wornoayporn, V.; Caceres, C.

2007-01-01

Irradiation of pupae in sterile insect technique (SIT) projects is usually undertaken in hypoxic atmospheres, which have been shown to lessen the deleterious effects of irradiation on the quality of adult sterile flies. Although this is the accepted technology in most mass-rearing and sterilization facilities, to date no information has been generated on the actual levels of oxygen (O 2 ) in pupae-packing containers during irradiation. The present study utilized recently-developed technology to investigate the O 2 level inside bags in which pupae of Mediterranean fruit fly (medfly) Ceratitis capitata (Wiedemann) are packed prior to irradiation, the ability of pupae to create hypoxic environments in these bags, and the effect of O 2 atmospheres on the quality of irradiated males. Pupae, 1 d before adult emergence, were shown to deplete the O 2 level in sealed bags in approximately 1 h. The rate of O 2 consumption was dependent upon pupal age and incubation temperature. Incubation temperature did not significantly affect the quality of pupae or mating capacity of resultant adult males if pupae were irradiated under maximal hypoxic conditions inside packing bags. In contrast, mating competitiveness drastically decreased when pupae were irradiated under ambient O 2 conditions, with the packing bag open. There was no difference in the mating capacity of males when pupae were irradiated in sealed bags under either 10% or 2% O 2 levels, or under maximal hypoxia. Normal doses of fluorescent dye, applied to pupae to mark sterile flies, did not affect the ability of pupae to create hypoxic conditions inside packing bags, nor the quality control parameters of either pupae or adults. Current practices in mass-rearing facilities are discussed in the light of these results. (author) [es

MISSE 6 Polymer Film Tensile Experiment

Science.gov (United States)

Miller, Sharon K. R.; Dever, Joyce A.; Banks, Bruce A.; Waters, Deborah L.; Sechkar, Edward; Kline, Sara

2010-01-01

The Polymer Film Tensile Experiment (PFTE) was flown as part of Materials International Space Station Experiment 6 (MISSE 6). The purpose of the experiment was to expose a variety of polymer films to the low Earth orbital environment under both relaxed and tension conditions. The polymers selected are those commonly used for spacecraft thermal control and those under consideration for use in spacecraft applications such as sunshields, solar sails, and inflatable and deployable structures. The dog-bone shaped samples of polymers that were flown were exposed on both the side of the MISSE 6 Passive Experiment Container (PEC) that was facing into the ram direction (receiving atomic oxygen, ultraviolet (UV) radiation, ionizing radiation, and thermal cycling) and the wake facing side (which was supposed to have experienced predominantly the same environmental effects except for atomic oxygen which was present due to reorientation of the International Space Station). A few of the tensile samples were coated with vapor deposited aluminum on the back and wired to determine the point in the flight when the tensile sample broke as recorded by a change in voltage that was stored on battery powered data loggers for post flight retrieval and analysis. The data returned on the data loggers was not usable. However, post retrieval observation and analysis of the samples was performed. This paper describes the preliminary analysis and observations of the polymers exposed on the MISSE 6 PFTE.

Cycloadditions of ketene acetals under microwave irradiation in solvent-free conditions

International Nuclear Information System (INIS)

Diaz-Ortiz, A.; Diez-Barra, E.; La Hoz, A. De; Prieto, P.; Moreno, A.

1994-01-01

When subjected to microwave irradiation ketene acetals undergo 1.3-dipolar and hetero-Diels-Alder cycloadditions within 5-12 min to give excellent yields of easily purified heterocyclic products. This efficient and rapid synthesis has the advantage of employing milder reaction conditions than those of classical thermal heating. (author)

Gamma radiation processed bamboo polymer composites. III. Possible applications for tensile reinforcement of concrete

International Nuclear Information System (INIS)

Adur, A.M.

1978-01-01

Three species of bamboo were converted to bamboo-polymer composites by vacuum impregnation with monomer and in situ polymerization using gamma irradiation. Resistance of the composites to various chemicals present in concrete was tested. Resistance to termites, fungus and other forms of biological attack was examined. Strength-to-weight ratios were calculated based on mechanical tests performed earlier (paper II of this three-part series). Possible application for tensile reinforcement of concrete is discussed in considerable detail. 2 figures, 4 tables

Fractography evaluation of impact and tensile specimens from the HFBR [High Flux Beam Reactor

International Nuclear Information System (INIS)

Czajkowski, C.J.

1989-10-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) has performed a fractographic examination of neutron irradiated and unirradiated tensile and Charpy ''V'' notch specimens. The evaluation was carried out using a scanning electron microscope (SEM) to evaluate the fracture mode. Photomicrographs were then evaluated to determine the extent of ductility present on the fracture surfaces of the unirradiated specimens. Ductility area measurements ranged from 4.6--9.5% on typical photomicrographs examined. 12 figs

Hydrogel Based on Crosslinked Methylcellulose Prepared by Electron Beam Irradiation for Wound Dressing Application

Directory of Open Access Journals (Sweden)

Ambyah Suliwarno

2014-10-01

Full Text Available The aim of this research is to explore the possibility of methylcellulose polymer to be used as wound dressing material prepared using electron beam technique. The methylcellulose paste solution with various of molecular weight (SM-4, SM-100, SM-400, SM-4000 and SM-8000 at different concentration (15-30% w/v were irradiated by using electron beam on the dose range of 10 kGy up to 40 kGy. Gel fraction and swelling ratio of hydrogels were determined gravimetrically. Tensile strength and elasticity of hydrogels were measured using a universal testing machine. It was found that with the increasing of irradiation dose from 10 up to 40 kGy, gel fraction and tensile strength were increased for all of hydrogels with various of molecular weight. On contrary, the swelling ratio of hydrogels decreased with increasing of irradiation dose. The optimum hydrogels elasticity were obtained from methylcellulose solution with the concentration range of 15-20% with irradiation dose of 20 kGy and showed excellent performance. The hydrogels based on methylcellulose prepared by electron beam irradiation can be considered for wound dressing material.

Trial production of low protein irradiated natural rubber latex by low energy electron beam in pilot scale

International Nuclear Information System (INIS)

Utama, Marga; Yoshii, F.; Kume, T.

2006-01-01

Three importance factors for producing low protein by low energy electron beam (250 keV/10 mA) irradiation in pilot scale (20 liters per bath) with 1,9-nonediol diacrylate (NDA) namely: maturation time of natural rubber latex before irradiation, treatment of irradiated natural rubber latex (INRL) before and after centrifugation, and standard irradiation method has been carried out. The results showed that the optimum irradiation time for producing INRL with 5 phr (part hundred ratio of rubber) of NDA as sensitize agent, and with the rotation speed of agitation 210 rpm (rotation per minutes) was between 20-30 minutes. By using this condition tensile strength of the INRL film was 26 MPa. The maturation of natural rubber latex before irradiation is the key for driving the quality of INRL. Water extractable protein content of INRL after leaching in 1% ammonia solution for 30 minutes at room temperature was around 47 μ/g, and after adding with 1 phr of PVA (poly vinyl alcohol) or 0.1 phr CMC (carboxy methyl cellulose) the water extractable protein content decrease less than 6 μ/g. (author)

Tensile properties and fracture of (α+γ) two phase stainless steel with fine grained microstructure

International Nuclear Information System (INIS)

Ogiyama, Hiroyuki; Tsukuda, Hitoshi; Soyama, Yoshiro

1989-01-01

The tensile properties and fracture of the (α+γ) two phase stainless steel with very fine γ grains were investigated. Two different microstructures with very fine γ grains can be obtianed by the thermomechanical treatments; One has both very fine γ and α grains, and the other has very fine γ and large α grains. The specimens were prepared in quenched and aged (475degC) conditions. The results obtained are as follows. The 0.2 % proof stress and tensile strength increase with the aging at 475degC for all specimens. The refinement of the γ grains plays an important role for the increase of strength in both quenched and aged conditions, and also for the ductility in the quenched condition. The refinement of α grains, however, contributes to the increase of the ductility in the aged condition. Accordingly, it is found that very good combinations of tensile strength and ductility can be achieved by the aging and the refinement of the γ and α grains (micro duplex structure). (author)

Pristine and γ-irradiated halloysite reinforced epoxy nanocomposites - Insight study

Science.gov (United States)

Saif, Muhammad Jawwad; Naveed, Muhammad; Zia, Khalid Mahmood; Asif, Muhammad

2016-10-01

The present study focuses on development of epoxy system reinforced with naturally occurring halloysite nanotubes (HNTs). A comparative study is presented describing the performance of pristine and γ-irradiated HNTs in an epoxy matrix. The γ-irradiation treatment was used for structural modification of natural pristine HNTs under air sealed environment at different absorbed doses and subsequently these irradiated HNTs were incorporated in epoxy resin with various wt% loadings. The consequences of γ-irradiation on HNTs were studied by FTIR and X-ray diffraction analysis (XRD) in terms of changes in functional groups and crystalline characteristics. An improvement is observed in mechanical properties and crack resistance of composites reinforced with γ-irradiated HNTs. The irradiated HNTs imparted an improved flexural and tensile strength/modulus along with better thermal performance.

Unstable propagation behavior of a ductile crack in SUS-304 stainless steel under high compliance tensile loading

International Nuclear Information System (INIS)

Tomoda, Yoshio

1981-01-01

In relation to the safe maintenance of nuclear power plants, it is necessary to prevent reactor coolant pipings from burst type failure caused by the unstable propagation of defects and cracks, such as stress corrosion cracking and fatigue cracks. In ductile materials, crack propagation is stable in tensile loading under fixed grip condition, when a specimen is controlled to deform in proportion to the increase of tensile load. However, it has been known that the instability of ductile cracks occurs after tensile load reached the maximum, especially under constant loading condition arising in the loading devices with high compliance or low tensile rigidity. In order to confirm the reliability of SUS 304 stainless pipes subjected to SCC, the crack propagation behavior was examined with the specimens having center cracks, using both testing machines with high compliance and low compliance. The instability of ductile cracks and the propagation velocity of unstable cracks were analyzed, and the calculated results were compated with the experimental results. Not only the compliance of testing machines but also the conditions of specimens affected the propagation of cracks. (Kako, I.)

The study on the threshold strain of microvoid formation in TRIP steels during tensile deformation

International Nuclear Information System (INIS)

Wang Wurong; Guo Bimeng; Ji Yurong; He Changwei; Wei Xicheng

2012-01-01

Highlights: ► The tensile mechanical behaviors of TRIP steels were studied under high rate deformation conditions. ► The threshold strain of microvoid formation was examined quantitatively. ► The effects of retained austenite of TRIP on suppressing microvoid formed during tensile process have been discussed. - Abstract: Transformation Induced Plasticity (TRIP) steels exhibit a better combination of strength and ductility properties than conventional high strength low alloy (HSLA) steels, and therefore receive considerable attention in the automotive industry. In this work, the tensile mechanical behaviors of TRIP-aided steels were studied under the condition of the quasi-static and high deformed rates. The deformed specimens were observed by scanning electron microscope (SEM) along the tensile axis. The threshold strain of microvoid formation was examined quantitatively according to the evolution of deformation. The results showed that: the yield and tensile strengths of TRIP steels increase with the strain rate, whereas their elongations decrease. However, the threshold strain for TRIP steels at high strain rate is larger than that at low strain rate. Comparing with the deformed microstructure and microvoids formed in the necking zone of dual phase (DP) steel, the progressive deformation-induced transformation of retained austenite in TRIP steels remarkably increases the threshold strain of microvoid formation and furthermore postpones its growth and coalescence.

Dislocation sweeping of defects in neutron- and electron-irradiated niobium

International Nuclear Information System (INIS)

Loomis, B.A.; Otero, M.P.

1983-10-01

The glide of dislocations in a [441]-oriented Nb single crystal irradiated at 325 K with 5.5 x 10 21 neutrons/m 2 (E > 0.1 MeV) is shown for increasing time of tensile elongation (2 x 10 -4 mm/s) in the High Voltage Electron Microscope at Argonne National Laboratory. The dimensions of the tensile specimen in the guage length were approximately 2 mm x 0.5 mm x 0.0001 mm. An electron energy of 900 keV was used during the simultaneous deformation and TEM observation

A Study of Tensile Flow and Work-Hardening Behavior of Alloy 617

Science.gov (United States)

Singh, Aditya Narayan; Moitra, A.; Bhaskar, Pragna; Dasgupta, Arup; Sasikala, G.; Bhaduri, A. K.

2018-04-01

The simple power relationship σ = Κɛ p n satisfactorily expresses the tensile flow behavior of many metals and alloys in their uniform plastic strain regime. However, many FCC materials with low stacking fault energy have opposed such power law relationship. Alloy 617, an age-hardenable Ni-based superalloy is also observed not to obey the simple power law relationship neither in its solution-treated nor in its aged conditions. Various flow relationships were used to obtain the best fit for the tensile data, and different relationships were identified for the different aged conditions. The work-hardening rate (θ) demonstrates three distinct regions for all aged conditions, and there is an obvious change in the trend of θ versus σ. In the initial portion, θ decreases rapidly followed by a gradual increase in the second stage and again a decrease in its third stage is perceived in the Alloy 617. These three-stage characteristics are attributed to a commonly known precipitate, γ': Ni3(Ti, Al) which evolves during aging treatment and well recognized under transmission electron microscopy (TEM) observation. TEM results also reveal a slight degree of coarsening in γ' over aging. The tensile flow and the work-hardening behavior are well correlated with other microstructural evolution during the aging treatments.

Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N.; Edwards, D.J.; Bilde-Soerensen, J.B

2004-10-01

The phenomenon of plastic flow localization in the form of 'cleared' channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels are initiated during post-irradiation deformation has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons. Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron irradiated at 323 and 373K to displacement doses in the range of 0.01 to 0.3 dpa (displacement per atom) and tensile tested at the irradiation temperature. The stress-strain curves clearly indicated the occurrence of a yield drop. The post-deformation microstructural examinations revealed that the channels are formed already in the elastic regime and their density increases with increasing plastic strain. The channels appear to have been initiated at grain boundaries, twin boundaries, at relatively large inclusions and even at the previously formed cleared channels. Even though the channels are produced throughout the whole tensile test, no clear evidence has been found for the operation of Frank-Read sources in the volume between the channels. Channels have been observed to penetrate through annealing twins, in some cases stopping at the opposite twin boundary and in other cases penetrating even through the opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimens activate dislocation sources at the sites of stress concentration at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications

Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

International Nuclear Information System (INIS)

Singh, B.N.; Edwards, D.J.; Bilde-Soerensen, J.B.

2004-10-01

The phenomenon of plastic flow localization in the form of 'cleared' channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels are initiated during post-irradiation deformation has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons. Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron irradiated at 323 and 373K to displacement doses in the range of 0.01 to 0.3 dpa (displacement per atom) and tensile tested at the irradiation temperature. The stress-strain curves clearly indicated the occurrence of a yield drop. The post-deformation microstructural examinations revealed that the channels are formed already in the elastic regime and their density increases with increasing plastic strain. The channels appear to have been initiated at grain boundaries, twin boundaries, at relatively large inclusions and even at the previously formed cleared channels. Even though the channels are produced throughout the whole tensile test, no clear evidence has been found for the operation of Frank-Read sources in the volume between the channels. Channels have been observed to penetrate through annealing twins, in some cases stopping at the opposite twin boundary and in other cases penetrating even through the opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimens activate dislocation sources at the sites of stress concentration at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these

An inverse method based on finite element model to derive the plastic flow properties from non-standard tensile specimens of Eurofer97 steel

Directory of Open Access Journals (Sweden)

S. Knitel

2016-12-01

Full Text Available A new inverse method was developed to derive the plastic flow properties of non-standard disk tensile specimens, which were so designed to fit irradiation rods used for spallation irradiations in SINQ (Schweizer Spallations Neutronen Quelle target at Paul Scherrer Institute. The inverse method, which makes use of MATLAB and the finite element code ABAQUS, is based upon the reconstruction of the load-displacement curve by a succession of connected small linear segments. To do so, the experimental engineering stress/strain curve is divided into an elastic and a plastic section, and the plastic section is further divided into small segments. Each segment is then used to determine an associated pair of true stress/plastic strain values, representing the constitutive behavior. The main advantage of the method is that it does not rely on a hypothetic analytical expression of the constitutive behavior. To account for the stress/strain gradients that develop in the non-standard specimen, the stress and strain were weighted over the volume of the deforming elements. The method was validated with tensile tests carried out at room temperature on non-standard flat disk tensile specimens as well as on standard cylindrical specimens made of the reduced-activation tempered martensitic steel Eurofer97. While both specimen geometries presented a significant difference in terms of deformation localization during necking, the same true stress/strain curve was deduced from the inverse method. The potential and usefulness of the inverse method is outlined for irradiated materials that suffer from a large uniform elongation reduction.

The effect of irradiation and hydration upon the mechanical properties of tendon

International Nuclear Information System (INIS)

Smith, C.W.; Kearney, J.N.

1996-01-01

Irradiation sterilization is in wide use among tissue banks, for both hard and soft tissue grafts. Irradiation of tendon can impair its mechanical properties. Following implantation of a tendon graft, re-vascularization and resorption processes reduce its mechanical performance. Tendon with severely impaired properties may not be suitable for use as a load-bearing graft, e.g. as anterior cruciate ligament replacement. An important factor determining the extent of the reduction of the mechanical performance is the condition of the tendon during irradiation, especially the presence of water. There has not yet been a study of the effects of both irradiation dose and hydration on tendon mechanical properties. This study measured the changes in tensile mechanical properties, including strength and stiffness, following γ irradiation doses of 15 kGy (1.5 MRad) and 25 kGy irradiated tendons was lower compared to fresh tendons, whereas the strength of the frozen irradiated tendons was very similar to that of the fresh. The tangent modulus of both of the freeze-dried irradiated groups were lower than the fresh tendons, as was the 15 kGy frozen group. The modulus of the 25 kGy frozen irradiated group was similar to the fresh. The general pattern of the results indicate that the two freeze-dried tendon groups were more affected than the frozen irradiated, and of the frozen irradiated groups the 25 kGy group was least affected. The results fit well with suggested mechanisms for the action of irradiation upon collagen; that intramolecular crosslinking and scission of the tropocollagen α chains occur when water is present, and α chain scission alone occurs when water is absent. Irradiation of tendons for use as grafts may produce minimal deleterious changes if the irradiation is performed while the tendon is frozen with water present. (Author)

Enhancement in irradiated mononuclear cells in culture of mitogen-induced incorporation of [3H]thymidine by homologous conditioned medium

International Nuclear Information System (INIS)

Sandru, G.; Greiner, R.

1994-01-01

Incorporation of [ 3 H]thymidine in irradiated peripheral blood mononuclear cell cultures irradiated in vitro was stimulated significantly by either concanavalin A or phytohemagglutinin only in the presence of homologous conditioned medium. Production of this activity by mononuclear cells was enhanced by irradiation and/or pulsed exposure to puromycin but was abolished by actinomycin D. Addition of anti-interleukin 1 or anti-interleukin 2 monoclonal antibodies to the conditioned medium before assay did not influence the stimulatory action. A similar significant stimulation of mononuclear cell cultures irradiated with 6 Gy by concanavalin A was obtained when purified preparations of homologous conditioned medium were used in the assay. Purification was done by ultrafiltration and concentration, heparin agarose chromatography, ammonium sulfate precipitation, concanavalin A agarose chromatography, DEAE-ion exchange chromatography and HPLC gel filtration chromatography. With SDS-PAGE and silver staining, the active HPLC fraction gave one band of 50 kDa, suggesting that this protein is responsible for the co-stimulatory effect of homologous conditioned medium for both mitogen-induced irradiated and nonirradiated mononuclear cell cultures. 42 refs., 9 figs., 3 tabs

Effects of neutron irradiation on mechanical properties of silicon carbide composites fabricated by nano-infiltration and transient eutectic-phase process

International Nuclear Information System (INIS)

Koyanagi, T.; Hinoki, T.; Shimoda, K.; Ozawa, K.; Katoh, Y.

2014-01-01

Unidirectional silicon carbide (SiC)-fiber-reinforced SiC matrix (SiC/SiC) composites fabricated by a nano-infiltration and transient eutectic-phase (NITE) process were irradiated with neutrons at 830°C to 5.9 dpa, and at 1270°C to 5.8 dpa. The in-plane and trans-thickness tensile and the inter-laminar shear properties were evaluated at ambient temperature. The mechanical characteristics, including the quasi-ductile behavior, the proportional limit stress, and the ultimate tensile strength, were retained subsequent to irradiation. Analysis of the stress–strain hysteresis loop indicated the increased fiber/matrix interface friction and the decreased residual stresses. The inter-laminar shear strength exhibited a significant decrease following irradiation. (author)

Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

International Nuclear Information System (INIS)

Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

1991-01-01

The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

Prevention Effect of Poly-Gamma-Glutamic Acid on Tissue Damage Induced by Gamma Irradiation as a Natural Cross-Linker

Energy Technology Data Exchange (ETDEWEB)

Kim, Jaehun; Sung, Nakyun; Kim, Jeongsoo; Jo, Euri; Choi, Jongil; Park, Jongheum; Lee, Juwoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Kwangwon [Eulji Univ. Hospital, Daejeon (Korea, Republic of); Kwon, Jungkee [Chonbuk National Univ., Jeonju (Korea, Republic of); Kim, Taewoon [Jeonbuk Technopark, Jeonju (Korea, Republic of)

2012-03-15

This study was to determine the prevention effect of poly-gamma-glutamic acid (PGA) on tissue damage induced by gamma irradiation for development of xenograft. PGA (MW 2000 kDa) extracted from permeated soy bean (natto) was used in this study as natural compound, and glutaraldehyde (GA) and 1-ethyl-3-(3-dimethylaminopropyl) carbodiimide (EDC) were used as a control, chemical based cross-linking agents. GA, EDC and PGA treated porcine tendons were gamma-irradiated at the dose of 30 kGy. Prevention effects against tissue damage were measured as the result of tensile strength, hydroxyproline contents and tissue morphological analysis. Tensile of porcine tendon was remarkably decreased by gamma irradiation, but increased in PGA treated group. Morphological analysis showed that collagen structure was broken by gamma irradiation, but attenuated by PGA treatment. Base on the results, it demonstrated that gamma irradiation can induce severe alteration of porcine tendon, but PGA can effectively improve the tissue damage.

Prevention Effect of Poly-Gamma-Glutamic Acid on Tissue Damage Induced by Gamma Irradiation as a Natural Cross-Linker

International Nuclear Information System (INIS)

Kim, Jaehun; Sung, Nakyun; Kim, Jeongsoo; Jo, Euri; Choi, Jongil; Park, Jongheum; Lee, Juwoon; Lee, Kwangwon; Kwon, Jungkee; Kim, Taewoon

2012-01-01

This study was to determine the prevention effect of poly-gamma-glutamic acid (PGA) on tissue damage induced by gamma irradiation for development of xenograft. PGA (MW 2000 kDa) extracted from permeated soy bean (natto) was used in this study as natural compound, and glutaraldehyde (GA) and 1-ethyl-3-(3-dimethylaminopropyl) carbodiimide (EDC) were used as a control, chemical based cross-linking agents. GA, EDC and PGA treated porcine tendons were gamma-irradiated at the dose of 30 kGy. Prevention effects against tissue damage were measured as the result of tensile strength, hydroxyproline contents and tissue morphological analysis. Tensile of porcine tendon was remarkably decreased by gamma irradiation, but increased in PGA treated group. Morphological analysis showed that collagen structure was broken by gamma irradiation, but attenuated by PGA treatment. Base on the results, it demonstrated that gamma irradiation can induce severe alteration of porcine tendon, but PGA can effectively improve the tissue damage

The application of an internal state variable model to the viscoplastic behavior of irradiated ASTM 304L stainless steel

Energy Technology Data Exchange (ETDEWEB)

McAnulty, Michael J., E-mail: mcanulmj@id.doe.gov [Department of Energy, 1955 Fremont Avenue, Idaho Falls, ID 83402 (United States); Potirniche, Gabriel P. [Mechanical Engineering Department, University of Idaho, Moscow, ID 83844 (United States); Tokuhiro, Akira [Mechanical Engineering Department, University of Idaho, Idaho Falls, ID 83402 (United States)

2012-09-15

Highlights: Black-Right-Pointing-Pointer An internal state variable approach is used to predict the plastic behavior of irradiated metals. Black-Right-Pointing-Pointer The model predicts uniaxial tensile test data for irradiated 304L stainless steel. Black-Right-Pointing-Pointer The model is implemented as a user-defined material subroutine in the finite element code ABAQUS. Black-Right-Pointing-Pointer Results are compared for the unirradiated and irradiated specimens loaded in uniaxial tension. - Abstract: Neutron irradiation of metals results in decreased fracture toughness, decreased ductility, increased yield strength and increased ductile-to-brittle transition temperature. Designers use the most limiting material properties throughout the reactor vessel lifetime to determine acceptable safety margins. To reduce analysis conservatism, a new model is proposed based on an internal state variable approach for the plastic behavior of unirradiated ductile materials to support its use for analyzing irradiated materials. The proposed modeling addresses low temperature irradiation of 304L stainless steel, and predicts uniaxial tensile test data of irradiated experimental specimens. The model was implemented as a user-defined material subroutine (UMAT) in the finite element software ABAQUS. Results are compared between the unirradiated and irradiated specimens subjected to tension tests.

Microstructure-tensile property correlation of 316SS in low dose neutron irradiations

International Nuclear Information System (INIS)

Yoshida, N.; Muroga, T.; Araki, K.; Heinisch, H.L.; Kiritani, M.

1990-05-01

The objective of this work is to determine the effects of the neutron spectrum on radiation-induced changes in mechanical properties for metals irradiated with fission and fusion neutrons. 10 refs., 6 figs

Heavy-section steel irradiation program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

Gamma irradiation effect on thermo shrink polyethylene tube: Case study

International Nuclear Information System (INIS)

Naurah Mat Isa; Siti Zulaiha Hairaldin; Mohd Yusof Hamzah; Zulkafli Ghazali

2010-01-01

The production technology of heat shrinkable tubes is based on the industrial usage of so-called memory effect in some polymers. Many polymers can be rendered heat shrinkable, particularly polyolefins by introducing crosslinking in the polymer backbone. Heat shrinkable tubes and moulded parts provide mechanical and/or chemical and/or electrical protection. . In this work, irradiation from Co-60 was used to impart crosslinking in polymer and thus modify their mechanical and thermal properties. Heat shrinkable or thermo shrink polyethylene (PE) tube with diameter 6.4 mm was selected to evaluate crosslinking behaviour from the gamma irradiation exposure at 5 different doses (120, 140, 160, 200 and 220 kGy). The gel content (%), tensile properties and thermal behavior and stability of the tubes were measured using DSC and TGA. Tubes irradiated at 140 kGy and 160 kGy showed better tensile properties than those at 120, 200 and 220 kGy doses. Maximum percent of crystallization was observed at 160 kGy and temperature required to crystallize is lower than melting temperature due to changes in molecular orientation. Thermal decomposition of the irradiated tubes occurred at 430-450 degree Celsius which is lower than the unirradiated sample. Gel content analysis for samples irradiated at 180-220 kGy yield up to 30 to 40 % gel fraction. Although the gel fractions are higher than other dose, the samples are expected to experience chain scission presumably due to localized gel formation. (author)

Irradiation effects in strain aged pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-02-15

Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

Swedish studies on irradiation effect in structural materials

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-12-15

A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed.

Swedish studies on irradiation effect in structural materials

International Nuclear Information System (INIS)

Grounes, M.; Myers, H.P.

1962-12-01

A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed

Experimental study associated to irradiation of FBR structural material, (4)

International Nuclear Information System (INIS)

1976-01-01

The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

Effect of surface treatment of titanium posts on the tensile bond strength

NARCIS (Netherlands)

Schmage, P; Sohn, J; Ozcan, M; Nergiz, [No Value

Objectives. Retention of composite resins to metal can be improved when metal surfaces are conditioned. The purpose of this investigation was to investigate the effect of two conditioning treatments on the tensile bond strength of four resin-based luting cements and zinc phosphate cement to titanium

Study on irradiation conditions of producing 153Sm with natural abundance samarium target

International Nuclear Information System (INIS)

Du Jin; Jin Xiaohai; Bai Hongsheng; Liu Yuemin; Chen Daming; Wang Fan

1998-01-01

Irradiation conditions of natural abundance 152 Sm targets in different forms are studied in the heavy water reactor and the light water swimming pool reactor at the China Institute of Atomic Energy. The result shows that the specific activity of 153 Sm in liquid form target irradiated in the light water swimming pool reactor is two times greater than that in solid form target. The radionuclide purity of 153 Sm is more than 99%, which can meet the needs of clinical application

Heavy-section steel irradiation program summary

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

1992-01-01

Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict failure models of the vessel material. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness (K Ic and J Ic ), crack-arrest toughness (K Ia ), ductile tearing resistance (dJ/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. (orig./GL)

Tensile and Flexural Test on Kenaf Hybrid Composites

Science.gov (United States)

Salleh, Z.; Yunus, S.; Masdek, N. R. N. M.; Taib, Y. M.; Azhar, I. I. S.; Hyie, K. M.

2018-03-01

The widely use of synthetic materials like carbon and fiberglass in various industries such as automotive and aircraft has lead to human health and environment problems. Therefore, the use of natural fibres such as kenaf has received higher attention as reinforcement. Kenaf or the scientific name is Hibiscus Cannabinus. L is one of the group of Malvecea plant which in the early days, the application of kenaf served only rope and canvas. However, it has more advantages than synthetic materials such as; widely availaible, renewable, lightweight, non-abbrasiveness during processing, high specific strength, free from health hazard and biodegradeable. This study was carried out to investigate the effects of different arrangement of kenaf and fiberglass composites on Young’s Modulus. The material composite was hardened with polyester resin and their properties was characterized. The tensile and the flexural properties is determined using an Instron universal tensile testing machine and carried out by following ASTM D3039 for tensile and ASTM D790 for a flexural test. The experimental program was designed to correlate the flexural and tensile Young’s Modulus of kenaf and fiberglass composite under the same load condition but different arrangement of kenaf and fiberglass on the mold . The resistance to change in shape was described by the behavior and characteristic of the composite materials. The stiffness or the elastic modulus of the composite material was determined at the end of the experiment. The results obtained show that the [±90FG/0/90/90/0/±90FG] kenaf/fiberglass composite arrangement has the highest elastic value.

Polymerization of calcium caseinates solutions induced by gamma irradiation

International Nuclear Information System (INIS)

Lacroix, M.; Jobin, M.; Mezgheni, E.; Srour, M.; Boileau, S.

1998-01-01

Solutions of calcium caseinate (5%) combined with propylene glycol (PG) or triethylene glycol(TEG) (0, 2.5% and 5%) and used for the development of edible films and coatings, were irradiated at doses between 0 to 128 kGy. Solutions were chromatographed through toyopearl HW 55F resin to observe the effect of irradiation on cross-link reactions. In unirradiated calcium caseinate solutions, two peaks could be observed (fractions 30 and 37) while samples irradiated at 64 kGy and 128 kGy showed one shifted peak at fraction 32 and 29 respectively. No effect of the plasticizers was observed. According to proteins standards of knowed molecular weights, the molecular weight of calcium caseinate increased approximately 10 times when irradiated at 128 kGy and 5 times when irradiated at 64 kGy. The physico-chemical properties of bio-films prepared with the irradiated solutions, demonstrated that tensile strength at break increased with increase of irradiation dose. A maximum dose was obtained at 16 kGy

Gamma irradiation induced effects of butyl rubber based damping material

Science.gov (United States)

Chen, Hong-Bing; Wang, Pu-Cheng; Liu, Bo; Zhang, Feng-Shun; Ao, Yin-Yong

2018-04-01

The effects of gamma irradiation on the butyl rubber based damping material (BRP) at various doses in nitrogen were investigated in this study. The results show that irradiation leads to radiolysis of BRP, with extractives increasing from 14.9 ± 0.8% of control to 37.2 ± 1.2% of sample irradiated at 350 kGy, while the swelling ratio increasing from 294 ± 3% to 766 ± 4%. The further investigation of the extractives with FTIR shows that the newly generated extractives are organic compounds containing C-H and C˭C bonds, with molecular weight ranging from 26,500 to 46,300. SEM characterization shows smoother surface with holes disappearing with increasing absorbed doses, consistent with "softer" material because of radiolysis. Dynamic mechanical study of BRP show that tan δ first slightly then obviously increases with increasing absorbed dose, while storage modulus slightly decreases. The tensile testing shows that the tensile strength decreases while the elongation at break increases with increasing dose. The positron annihilation lifetime spectroscopy show no obvious relations between free volume parameters and the damping properties, indicating the complicated influencing factors of damping properties.

Neutron irradiation behavior of ITER candidate beryllium grades

Energy Technology Data Exchange (ETDEWEB)

Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

1998-01-01

Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

Effect of laser fluence on surface, structural and mechanical properties of Zr after irradiation in the ambient environment of oxygen

International Nuclear Information System (INIS)

Jelani, M.; Bashir, S.; Khaleeq-ur Rehman, M.; Ahamad, R.; Ul-Haq, F.; Yousaf, D.; Akram, M.; Afzal, N.; Umer Chaudhry, M.; Mahmood, K.; Hayat, A.; Ahmad, Sajjad

2013-01-01

The laser irradiation effects on surface, structural and mechanical properties of zirconium (Zr) have been investigated. For this purpose, Zr samples were irradiated with Excimer (KrF) laser (λ = 248 nm, τ = 18 ns, repetition rate ∼ 30 Hz). The irradiation was performed under the ambient environment of oxygen gas at filling pressure of 20 torr by varying laser fluences ranging from 3.8 to 5.1 cm -2 . The surface and structural modification of irradiated targets was investigated by scanning electron microscope (SEM) and X-ray diffractometer (XRD). In order to explore the mechanical properties of irradiated Zr, the tensile testing and Vickers micro hardness testing techniques were employed. SEM analysis reveals the grain growth on the irradiated Zr surfaces for all fluences. However, the largest sized grains are grown for the lowest fluence of 3.8 J/cm 2 . With increasing fluence from 4.3 to 5.1 J cm -2 , the compactness and density of grains increase whereas their size decreases. XRD analysis reveals the appearance of new phases of ZrO 2 and Zr 3 O. The variation in the peak intensity is observed to be anomalous whereas decreasing trend in the crystallite size and residual stresses has been observed with increasing fluence. Micro hardness analysis reveals the increasing trend in surface hardness with increasing fluence. The tensile testing exhibits the increasing trend of yield stress (YS), decreasing trend of percentage elongation and anomalous behaviour of ultimate tensile strength with increasing fluence. (authors)

Thermal shock testing of ceramics with pulsed laser irradiation

International Nuclear Information System (INIS)

Benz, R.; Naoumidis, A.; Nickel, H.

1986-04-01

Arguments are presented showing that the resistance to thermal stressing (''thermal shock'') under pulsed thermal energy deposition by various kinds of beam irradiations is approximately proportional to Φ a √tp, where Φ a is the absorbed power density and tp is the pulse length, under conditions of diffusivity controlled spreading of heat. In practical beam irradiation testing, incident power density, Φ, is reported. To evaluate the usefulness of Φ√tp as an approximation to Φ a √tp, damage threshold values are reviewed for different kinds of beams (electron, proton, and laser) for a range of tp values 5x10 -6 to 2 s. Ruby laser beam irradiation tests were made on the following ceramics: AlN, BN, graphite, αSiC, β-SiC coated graphites, (α+β)Si 3 N 4 , CVD (chemical vapor deposition) TiC coated graphite, CVD TiC coated Mo, and CVD TiN coated IN 625. The identified failure mechanisms are: 1. plastic flow followed by tensile and bend fracturing, 2. chemical decomposition, 3. melting, and 4. loss by thermal spallation. In view of the theoretical approximations and the neglect of reflection losses there is reasonable accord between the damage threshold Φ√tp values from the laser, electron, and proton beam tests. (orig./IHOE)

Effect of gamma irradiation conditions on the radiation-induced degradation of isobutylene-isoprene rubber

Energy Technology Data Exchange (ETDEWEB)

Sen, M. E-mail: msen@hacettepe.edu.tr; Uzun, C.; Kantoglu, Oe.; Erdogan, S.M.; Deniz, V.; Gueven, O

2003-08-01

The effect of gamma irradiation conditions on the radiation-induced degradation of uncrosslinked, commercial isobutylene-isoprene rubbers has been investigated in this study. Influence of dose rate and irradiation atmosphere on the degradation of butyl rubber has been followed by viscosimetric and chromatographic analyses. Limiting viscosity number of all butyl rubbers decreased sharply up to 100 kGy and leveled off at around the same molecular weight, independent of dose rate. Slightly higher decrease in viscosity was observed for samples irradiated in air than in nitrogen especially at low dose rate irradiation. Cross-linking G(X), and chain scission G(S) yields of butyl rubbers were calculated by using weight- and number-average molecular weights of irradiated rubber determined by Size Exclusion Chromatography analyses. G-value results showed that chain scission reactions in isobutylene-isoprene rubber in air atmosphere are more favorable than in nitrogen atmosphere, and that lower dose rate enhances chain scission over cross-linking.

Tensile deformation and failure of North American porcupine quills

Energy Technology Data Exchange (ETDEWEB)

Chou, S.F.; Overfelt, R.A., E-mail: overfra@auburn.edu

2011-12-01

Although the mechanical properties of some keratin-based biological materials have been extensively studied (i.e., wool) and others are beginning to be studied (e.g., horn, hooves and avian quills), data on the properties of porcupine quill are less common. Porcupine quill is a keratin-based biological material composed of a cylindrical outer shell with an inner foam core. The present paper reports on the physical characteristics, tensile properties and fracture behavior of North American porcupine quills conditioned at relative humidities of 65% and 100%. Increasing the water content decreased the tensile stiffness and strength and increased the strain at fracture of the porcupine quills. The tensile fracture strength of porcupine quill was found to be 146 MPa at 65% RH and 60 MPa at 100% RH. Although these values compare favorably with reported values for African porcupine quill, reported values of the tensile strengths of wool with similar moisture contents are considerably higher. The initial moduli of porcupine quill (2700 MPa at 65% RH and 1000 MPa at 100% RH) compare favorably to those reported for wool but are considerably less than previous reports for African porcupine quill. The engineering strains at fracture were measured as 25% at 65% RH and 49% at 100% RH and these values are also comparable to other keratin-based mammalian materials. Scanning electron microscopy of the fracture surfaces of porcupine quills revealed that the cylindrical outer shells of quills are composed of 2-3 layers with distinctly different fracture characteristics, especially when the samples contain 100% RH. The outer layer of the porcupine quill shell appears to resist the plasticizing effects of moisture and appears to exhibit considerably less ductility than the inner layers, perhaps due to the presence of hydrophobic lipids in the outer layer. Highlights: {yields} We characterize the tensile properties of north American porcupine quill. {yields} Elastic modulus, tensile

Optimization of irradiation conditions for determination of LD50 in pigs

International Nuclear Information System (INIS)

Prochazka, Z.; Hampl, J.; Sedlacek, M.; Rodak, L.

1975-01-01

Radiation LDsub(50/30) values were determined in 36 twelve-week-old pigs (with a mean body weight of 21 kg) exposed to whole-body X-ray irradiation on a revolvable table rotated at a rate of 2.5 rpm using the following conditions: 180 kV, 15 mA, focal distance 79 cm, HVT 0.9 mm Cu, dose rate 2.42 x 10 -3 to 2.68 x 10 -3 C kg -1 min -1 (9.4 to 10.4 R/min) depending upon the animal size. The coefficient of mean irradiation uniformity was 1.4. Under these conditions the LDsub(50/30) for pigs was found to be 5.89 x 10 -2 C kg -1 , (228.3 R) with the biological range of effectiveness being 5.22 x 10 -2 to 6.90 x 10 -2 C kg -1 (202.4 to 267.6 R) Further experiments on 77 pigs showed that the LD 50 determined in this study had actually the median lethal effect. (orig.) [de

Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

International Nuclear Information System (INIS)

Kano, F.; Tsuchiya, Y.; Oka, K.

2009-01-01

The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

Energy Technology Data Exchange (ETDEWEB)

Singh, B N; Edwards, D J; Horsewell, A; Toft, P

1995-09-01

The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

International Nuclear Information System (INIS)

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

1995-09-01

The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10 17 n/m 2 s (E > 1 MeV, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

International Nuclear Information System (INIS)

Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

1979-01-01

An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n.cm -2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg.mm -2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

International Nuclear Information System (INIS)

Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

1979-02-01

An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n/cm 2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg/mm 2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

Effect of gamma irradiation on the properties of natural rubber/styrene butadiene rubber blends

Directory of Open Access Journals (Sweden)

A.B. Moustafa

2016-09-01

Full Text Available Blends of natural rubber (NR with styrene butadiene rubber (SBR with varying ratios have been prepared. Vulcanization of the prepared blends has been induced by irradiation of gamma rays with varying doses up to 250 kGy. Mechanical properties, namely tensile strength, tensile modulus at 100% elongation, elongation at break have been followed up as a function of irradiation dose as well as blend composition. Physical properties, namely gel fraction and swelling number have been followed up using benzene as a solvent. Thermal measurements namely thermogravimetric analysis were carried out. The results indicated that the addition of NR has improved the properties of NR / SBR blends. Also NR/SBR blend is thermally stable than NR alone.

Neutron flux and annealing effects on irradiation hardening of RPV materials

Science.gov (United States)

Chaouadi, R.; Gérard, R.

2011-11-01

This paper aims to examine an eventual effect of neutron flux, sometimes referred to as dose rate effect, on irradiation hardening of a typical A533B reactor pressure vessel steel. Tensile tests on both low flux (reactor surveillance data) and high flux (BR2 reactor) were performed in a large fluence range. The obtained results indicate two features. First, the surveillance data exhibit a constant (˜90 MPa) higher yield strength than the high flux data. However, this difference cannot be explained from a flux effect but most probably from differences in the initial tensile properties. The hardening kinetic of both low and high flux is the same. Annealing at low temperature, 345 °C/40 h, to eventually reveal unstable matrix damage did not affect both BR2 and surveillance specimens. This is confirmed by other annealing experimental data including both tensile and hardness measurements and tensile data on A508 forging and weld. It is suggested that the absence of flux effect on the tensile properties while different radiation-induced microstructures can be attributed to thermal ageing effects.

Characterization of ion irradiation effects on the microstructure, hardness, deformation and crack initiation behavior of austenitic stainless steel:Heavy ions vs protons

Science.gov (United States)

Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

2018-04-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a complex phenomenon of degradation which can have a significant influence on maintenance time and cost of core internals of a Pressurized Water Reactor (PWR). Hence, it is an issue of concern, especially in the context of lifetime extension of PWRs. Proton irradiation is generally used as a representative alternative of neutron irradiation to improve the current understanding of the mechanisms involved in IASCC. This study assesses the possibility of using heavy ions irradiation to evaluate IASCC mechanisms by comparing the irradiation induced modifications (in microstructure and mechanical properties) and cracking susceptibility of SA 304 L after both type of irradiations: Fe irradiation at 450 °C and proton irradiation at 350 °C. Irradiation-induced defects are characterized and quantified along with nano-hardness measurements, showing a correlation between irradiation hardening and density of Frank loops that is well captured by Orowan's formula. Both irradiations (iron and proton) increase the susceptibility of SA 304 L to intergranular cracking on subjection to Constant Extension Rate Tensile tests (CERT) in simulated nominal PWR primary water environment at 340 °C. For these conditions, cracking susceptibility is found to be quantitatively similar for both irradiations, despite significant differences in hardening and degree of localization.

Failure criteria for low-temperature irradiated organic composite insulation systems

International Nuclear Information System (INIS)

Schutz, J.B.; Fabian, P.E.

1997-01-01

Composite insulation systems in fusion magnet applications are often subjected to conditions of combined through thickness compression and shear at varying stress ratios. Characterization of insulation system strength under these conditions requires unidirectional shear and compression tests, as well as biaxial shear/compression testing to define the systems failure envelope. An appropriate failure criteria would reduce the number of tests required to define this envelope, and would give designers a better estimate of material strengths at intermediate stress ratios. Biaxial shear/compression testing requires the use of metallic sandwich specimens, which are susceptible to activation during irradiation. A reliable failure criteria which eliminates the need for biaxial characterization would also reduce the difficulty and expense involved in radiation effects characterization while still providing a complete failure envelope. Several generalized failure criteria, taken from classical composites analysis, were reformulated for application in the shear/compression quadrant. The maximum stress and Tsai-Hill criteria were found to be inadequate to represent the biaxial failure envelope. The generalized Tsai-Wu tensor quadratic failure criterion was also examined. Application of this criterion requires through thickness tensile strength data. Through thickness tensile strengths of insulation systems have not been routinely characterized, but limited data are available. Utilizing unidirectional through thickness tension, compression, and interlaminar shear strength data, the Tsai-Wu failure criteria was found to agree well with biaxial shear/compression test data

Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

Self-ion Irradiation Damage of F/M and ODS steels

International Nuclear Information System (INIS)

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

2014-01-01

Oxide dispersion strengthened (ODS) ferritic steels are potential high-temperature materials that are stabilized by dispersed particles at elevated temperatures. These dispersed particles improve the tensile strength and creep rupture strength, they are expected to increase the operation temperature up to approximately 650 .deg. C and also enhance the energy efficiency of the fusion reactor. Some reports described that the nano-clusters are strongly resistant to coarsening by annealing up to 1000 .deg. C, and nanoclusters do not change after ion irradiation up to 0.7 dpa at 300 .deg. C. ODS steels will be inevitably exposed to neutron irradiation condition; the irradiation damages, creep and swelling are always great concern. The dispersed oxide particles are believed to determine the performance of the steel, even the radiation resistance. In this study, F/M and ODS model alloys of Korea Atomic Energy Research Institute (KAERI) were irradiated by Fe 3+ self-ion to emulate the neutron irradiation effect. In this study, Fe 3+ self-ion irradiation is used as means of introducing radiation damage in F/M steel and ODS steel. The ion accelerator named DuET (in Kyoto University, Japan) was used for irradiation of Fe 3+ ion by 6.4 MeV at 300 .deg. C. The maximum damage rate in F/M and ODS steels were estimated roughly 6 dpa. After radiation, point or line defects were dominantly observed in F/M steel, on the other hands, small circular cavities were typically observed in ODS steel. Nanoindentation is a useful tool to determine the irradiationinduced hardness change in the damage layer of ionirradiated iron base alloys

Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

International Nuclear Information System (INIS)

Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

2000-03-01

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

Synthesis of fish scales gelatin-chitosan crosslinked films by gamma irradiation techniques

International Nuclear Information System (INIS)

Erizal; Perkasa, D.P.; Abbas, B.; Sulistioso, G.S.

2013-01-01

Gelatin is an important component of fish scales. Nowadays, attention has increased concerning the application of gelatin.The aim of this research was to improve the mechanical properties of gelatin produced from fish scales, which concurrently could increase the usefulness of fish scales. Gelatin (G) is prone to degrade or dissolve in water at room temperature, therefore to enhance its lifetime, it has to be modified with other compound such as chitosan. Chitosan (Cs) is a biodegradable polymer, which has biocompatibility and antibacterial properties. In this study, gelatin solution was mixed with chitosan solution in various ratios (G/Cs: 100/0, 75/25, 50/50, 25/75, 0/100), casted at room temperature to make composite films, then tested for the effectiveness of various gamma irradiation doses (10-40 kGy) for crosslinking of the two polymers. Chemical changes of the films were measured by FT-IR, gel fractions were determined by gravimetry, and mechanical properties were determined by tensile strength and elongation at break using universal testing machine. At optimum conditions ( 30 kGy and 75% Cs), the gel fraction, tensile strength, and elongation at break were higher leading to a stronger composite films as compared to the gelatin film. FTIR spectral analysis showed that gelatin and chitosan formed a crosslinked network. It was concluded that G-Cs films prepared by gamma irradiation have improved their mechanical properties than the gelatin itself. (author)

Insights into the effects of tensile and compressive loadings on human femur bone.

Science.gov (United States)

Havaldar, Raviraj; Pilli, S C; Putti, B B

2014-01-01

Fragile fractures are most likely manifestations of fatigue damage that develop under repetitive loading conditions. Numerous microcracks disperse throughout the bone with the tensile and compressive loads. In this study, tensile and compressive load tests are performed on specimens of both the genders within 19 to 83 years of age and the failure strength is estimated. Fifty five human femur cortical samples are tested. They are divided into various age groups ranging from 19-83 years. Mechanical tests are performed on an Instron 3366 universal testing machine, according to American Society for Testing and Materials International (ASTM) standards. The results show that stress induced in the bone tissue depends on age and gender. It is observed that both tensile and compression strengths reduces as age advances. Compressive strength is more than tensile strength in both the genders. The compression and tensile strength of human femur cortical bone is estimated for both male and female subjecting in the age group of 19-83 years. The fracture toughness increases till 35 years in male and 30 years in female and reduces there after. Mechanical properties of bone are age and gender dependent.

Tensile and creep data on type 316 stainless steel

Energy Technology Data Exchange (ETDEWEB)

Sikka, V. K.; Booker, B. L.P.; Booker, M. K.; McEnerney, J. W.

1980-01-01

This report summarizes tensile and creep data on 13 heats of type 316 stainless steel. It includes ten different product forms (three plates, four pipes, and three bars) of the reference heat tested at ORNL. Tensile data are presented in tabular form and analyzed as a function of temperature by the heat centering method. This method yielded a measure of variations within a single heat as well as among different heats. The upper and lower scatter bands developed by this method were wider at the lower temperatures than at the high temperatures (for strength properties), a trend reflected by the experimental data. The creep data on both unaged and aged specimens are presented in tabular form along with creep curves for each test. The rupture time data are compared with the ASME Code Case minimum curve at each test temperature in the range from 538 to 704{sup 0}C. The experimental rupture time data are also compared with the values predicted by using the rupture model based on elevated-temperature ultimate tensile strength. A creep ductility trend curve was developed on the basis of the reference heat data and those published in the literature on nitrogen effects. To characterize the data fully, information was also supplied on vendor, product form, fabrication method, material condition (mill-annealed vs laboratory annealed and aged), grain size, and chemical composition for various heats. Test procedures used for tensile and creep results are also discussed.

Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

International Nuclear Information System (INIS)

Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

2007-01-01

Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

Effect of γ-irradiation on the physical and mechanical properties of kefiran biopolymer film.

Science.gov (United States)

Shahabi-Ghahfarrokhi, Iman; Khodaiyan, Faramarz; Mousavi, Mohammad; Yousefi, Hossein

2015-03-01

In this study, the effect of different γ-ray dosages (3, 6, and 9 kGy) on the functional properties of kefiran biopolymer was investigated. The obtained results showed that increasing γ-ray dosage brought about an increase in the tensile strength of film specimens up to three-times. However, elongation at break, and tensile energy to break of γ-irradiated kefiran films decreased in the wake of increasing γ-ray dosage. γ-Irradiation could improve surface hydrophobicity, sensitivity of kefiran film specimens to water, and water vapor permeability, but yellowness of films increased, simultaneously. XRD spectrum confirmed increased crystallinity of γ-irradiated films. Melting point of films was constant but glass transition temperature decreased drastically at high γ-ray dosage (9 kGy). ATR-FTIR analysis confirmed that γ-ray engendered no changes in chemical functional groups. According to the result, a mechanism was proposed to percept the effects of γ-irradiation on kefiran biopolymer and its role on the functional properties of kefiran film. Hence, the functional properties of kefiran films were depend on the ratio of cross-linkages between polymer chains and produced mono and disaccharide by γ-irradiation. Copyright © 2014 Elsevier B.V. All rights reserved.

Effects of electron beam irradiation on properties of corn starch undergone periodate oxidation mechanism blended with polyvinyl alcohol

Science.gov (United States)

Bee, Soo-Tueen; Sin, Lee Tin; Ratnam, C. T.; Yap, Bee-Fen; Rahmat, A. R.

2018-02-01

This work was performed to examine the properties of pristine PVOH and PVOH-starch blends under exposure of different irradiation dosages. The periodate oxidation method was used to produce dialdehyde starch. The application of low dosages of electron beam irradiation (≤10 kGy) has improved the tensile strength by forming crosslinking networks. However, the tensile strength drastically declined when radiated at 30 kGy due to the reduction of available hydroxyl groups inside polymer matrix for intermolecular interaction. Also, the incorporation of corn starch and dialdehyde starch has significantly reduced the melting temperature and enthalpy of melting of PVOH blends due to cessation of the hydrogen bonding between PVOH and starch molecules. The crystallite size for deflection planes (1 0 1), (1 0 1) and (2 0 0) for all PVOH blends was significant reduced when irradiated. The electron beam irradiation has also weakened the hydrophilic characteristic of all PVOH blends as evidenced in infrared and microscopy analysis.

Storage tests with irradiated and non-irradiated onions

International Nuclear Information System (INIS)

Gruenewald, T.; Rumpf, G.; Troemel, I.; Bundesforschungsanstalt fuer Ernaehrung, Karlsruhe

1978-07-01

The results of several test series on the storage of irradiated and non-irradiated German grown onion are reported. Investigated was the influence of the irradiation conditions such as time and dose and of the storage conditions on sprouting, spoilage, browning of the vegetation centres, composition of the onions, strength and sensorial properties of seven different onion varieties. If the onions were irradiated during the dormancy period following harvest, a dose of 50 Gy (krad) was sufficient to prevent sprouting. Regarding the irradiated onions, it was not possible by variation of the storage conditions within the limits set by practical requirements to extend the dormancy period or to prevent browning of the vegetation centres, however. (orig.) 891 MG 892 RSW [de

Thermal analysis of polypropylene modified by gamma irradiation composites under outdoor conditions

International Nuclear Information System (INIS)

Komatsu, Luiz G.H.; Oliani, Washington L.; Lugao, Ademar B.; Parra, Duclerc F.

2015-01-01

This work reports the influence of the clay in the degradation process of the HMSPP.The polypropylene (PP) was irradiated under acetylene atmosphere in gamma irradiation source ( 60 Co) to obtain the HMSPP (high melt strength polypropylene). Composites of HMSPP were processed in twin-screw extruder with clay Cloisite 20A and Maleic Anhydride (PP-g-MA) as coupling agent. The obtained composites were exposed under outdoor conditions for 6 months. The ageing effects were characterized by Differential Scanning Calorimetry (DSC), Thermogravimetry Analysis (TGA). Chemical oxidation was evaluated by Carbonyl Index (IC) through infrared Spectroscopy (FT-IR). The results showed correlation between carbonyl index and ageing time. (author)

Thermal analysis of polypropylene modified by gamma irradiation composites under outdoor conditions

Energy Technology Data Exchange (ETDEWEB)

Komatsu, Luiz G.H.; Oliani, Washington L.; Lugao, Ademar B.; Parra, Duclerc F., E-mail: dfparra@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

This work reports the influence of the clay in the degradation process of the HMSPP.The polypropylene (PP) was irradiated under acetylene atmosphere in gamma irradiation source ({sup 60}Co) to obtain the HMSPP (high melt strength polypropylene). Composites of HMSPP were processed in twin-screw extruder with clay Cloisite 20A and Maleic Anhydride (PP-g-MA) as coupling agent. The obtained composites were exposed under outdoor conditions for 6 months. The ageing effects were characterized by Differential Scanning Calorimetry (DSC), Thermogravimetry Analysis (TGA). Chemical oxidation was evaluated by Carbonyl Index (IC) through infrared Spectroscopy (FT-IR). The results showed correlation between carbonyl index and ageing time. (author)

Overview and Critical Assessment of the Tensile Properties of unirradiated and irradiated EUROFER97

Energy Technology Data Exchange (ETDEWEB)

Lucon, E.; Vandermeulen, W.

2007-10-15

Material research represents a crucial issue for the assessment of fusion as a future viable source of energy. Structural materials, in particular, need to show a superior mechanical and chemical behaviour to guarantee the safe operation of the reactor during its whole lifetime, while retaining low activation characteristics to minimise the environmental impact of the produced waste. In this context, specific efforts have been focused for the last twenty years in Europe, Japan and the US, on developing suitable Reduced Activation Ferritic Martensitic (RAFM) steels as candidate structural materials. EUROFER97 has recently emerged in Europe as the reference material for the DEMO design. In the framework of the Long-Term Programme of EFDA (European Fusion Development Agreement), a coordinated effort has been launched aimed at providing a critical assessment of the mechanical and microstructural properties of EUROFER97 in the unirradiated and irradiated conditions, based on the results accumulated since the late 90ies within numerous EFDA tasks.

Irradiation of goods

International Nuclear Information System (INIS)

Huebner, G.

1992-01-01

The necessary dose and the dosage limits to be observed depend on the kind of product and the purpose of irradiation. Product density and density distribution, product dimensions, but also packaging, transport and storage conditions are specific parameters influencing the conditions of irradiation. The kind of irradiation plant - electron accelerator or gamma plant - , its capacity, transport system and geometric arrangement of the radiation field are factors influencing the irradiation conditions as well. This is exemplified by the irradiation of 3 different products, onions, deep-frozen chicken and high-protein feed. Feasibilities and limits of the irradiation technology are demonstrated. (orig.) [de

Pristine and γ-irradiated halloysite reinforced epoxy nanocomposites – Insight study

International Nuclear Information System (INIS)

Saif, Muhammad Jawwad; Naveed, Muhammad; Zia, Khalid Mahmood; Asif, Muhammad

2016-01-01

The present study focuses on development of epoxy system reinforced with naturally occurring halloysite nanotubes (HNTs). A comparative study is presented describing the performance of pristine and γ-irradiated HNTs in an epoxy matrix. The γ-irradiation treatment was used for structural modification of natural pristine HNTs under air sealed environment at different absorbed doses and subsequently these irradiated HNTs were incorporated in epoxy resin with various wt% loadings. The consequences of γ-irradiation on HNTs were studied by FTIR and X-ray diffraction analysis (XRD) in terms of changes in functional groups and crystalline characteristics. An improvement is observed in mechanical properties and crack resistance of composites reinforced with γ-irradiated HNTs. The irradiated HNTs imparted an improved flexural and tensile strength/modulus along with better thermal performance. - Highlights: • The γ-irradiation was used for structural modification of halloysite nanotubes. • Composite materials with irradiated HNTs showed improved mechanical properties. • The γ-irradiation treatment is a promising surface modification method.

Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U–Mo/Zr fuel plates under a heterogeneous irradiation condition

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yunmei; Gong, Xin; Ding, Shurong, E-mail: dsr1971@163.com

2015-04-15

Highlights: • The three-dimensional stress update algorithms in a co-rotational framework are developed for U–Mo and Zircalloy with the irradiation effects. • An effective method for three-dimensional modeling of the in-pile behaviors in heterogeneously irradiated monolithic fuel plates is established and validated. • The effects of the fission-induced creep effects in the U–Mo foil are investigated in detail. • A deformation phenomenon similar to the irradiation experimental results is obtained. - Abstract: For monolithic fuel plates with U–Mo foil and Zircalloy cladding, the three-dimensional large deformation incremental constitutive relations and stress update algorithms in the co-rotational coordinate framework are developed for the fuel and cladding with their respective irradiation effects involved. Three-dimensional finite element simulation of their in-pile thermo-mechanical coupling behaviors under a location-dependent irradiation condition is implemented via the validated user-defined subroutines UMATHT and UMAT in ABAQUS. Comparison of the simulation results for two cases with or without creep considered in the U–Mo foil indicates that with the irradiation creep included (1) considerable stress-relaxation appears in the U–Mo foil, and the mechanical interaction between fuel and cladding is weakened; (2) approximately identical thickness increments in the plate and fuel foil exist and become comparably larger; (3) plastic deformation in the cladding is significantly diminished.

Ratcheting Strain and Microstructure Evolution of AZ31B Magnesium Alloy under a Tensile-Tensile Cyclic Loading.

Science.gov (United States)

Yan, Zhifeng; Wang, Denghui; Wang, Wenxian; Zhou, Jun; He, Xiuli; Dong, Peng; Zhang, Hongxia; Sun, Liyong

2018-03-28

In this paper, studies were conducted to investigate the deformation behavior and microstructure change in a hot-rolled AZ31B magnesium alloy during a tensile-tensile cyclic loading. The relationship between ratcheting effect and microstructure change was discussed. The ratcheting effect in the material during current tensile-tensile fatigue loading exceeds the material's fatigue limit and the development of ratcheting strain in the material experienced three stages: initial sharp increase stage (Stage I); steady stage (Stage II); and final abrupt increase stage (Stage III). Microstructure changes in Stage I and Stage II are mainly caused by activation of basal slip system. The Extra Geometrically Necessary Dislocations (GNDs) were also calculated to discuss the relationship between the dislocation caused by the basal slip system and the ratcheting strain during the cyclic loading. In Stage III, both the basal slip and the {11-20} twins are found active during the crack propagation. The fatigue crack initiation in the AZ31B magnesium alloy is found due to the basal slip and the {11-20} tensile twins.

Low temperature gamma-ray irradiation effects on polymer materials

International Nuclear Information System (INIS)

Kudoh, Hisaaki; Kasai, Noboru; Sasuga, Tsuneo; Seguchi, Tadao

1995-01-01

The gamma radiation induced degradation of glass fiber reinforced plastic (GFRP) and polymethylmethacrylate (PMMA) at 77K was examined by flexural test and gas analysis after irradiation and compared by the irradiation at room temperature. The decrease in flexural strength at break was much less at 77K than at RT. The evolution of CH 4 , CO and CO 2 was also depressed at 77K. The temperature dependence of the degradation closely relates to the local molecular motion of matrix resin during irradiation. Polytetrafluoroethylene (PTFE) was also studied by irradiation at RT, 77K and 4K in terms of tensile elongation and molecular weight. The degradation was much less at 77K and 4K than at RT, and the same between 77K and 4K. (author)

Irradiation creep of candidate materials for advanced nuclear plants

Energy Technology Data Exchange (ETDEWEB)

Chen, J., E-mail: jiachao.chen@psi.ch; Jung, P.; Hoffelner, W.

2013-10-15

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

Electron irradiation effect on single crystal of niobium

International Nuclear Information System (INIS)

Otero, M.P.; Lucki, G.

1984-01-01

The effect of electron irradiation (900 KeV) on gliding dislocations of single crystal Nb with its tensile axe in the [941] orientation was observed for the in-situ deformation in a high voltage electron microscope (HVEM) at Argonne National Laboratory. The experimental was carried out by the 1 hour-electron irradiation with no stress applied. Straight dislocations actuating as sinks for the electron produced defects became helicoidal as the irradiation proceeded. Frenkel pairs were created in Nb for electron energies > = 650 KeV and, as the single vacancies do not undergo long-range migration in Nb at temperatures much below 620 K, the defects that are entrapped by the dislocations are self-interstitials produced by electron displacement. Applying the stress it was possible to observe that modified dislocations did not glide while the dislocations not affected by the irradiation are visibly in movement. This important result explains the neutron and electron-irradiation induced work-hardening effect for Nb that was previously observed. (Author) [pt

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Corwin, W.R.

1990-08-01

The primary goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (particularly the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program includes direct continuation of irradiation studies previously conducted by the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are examined on a wide range of fracture properties. Detailed statistical analyses of the fracture data on K Ic shift of high-copper welds were performed. Analysis of the first phase of irradiated crack-arrest testing on high-copper welds was completed. Final analysis and publication of the results of the second phase of the irradiation studies on stainless steel weld-overlay cladding were completed. Determinations were made of the variations in chemistry and unirradiated RT NDT of low upper-shelf weld metal from the Midland reactor. Final analyses were performed on the Charpy impact and tensile data from the Second and Third Irradiation series on low upper-shelf welds, and the report on the series was drafted. A detailed survey of existing data on microstructural models and data bases of irradiation damage was performed, and initial development of a reaction-rate-based model was completed. 40 refs., 7 figs., 4 tabs

Tensile Strength of Water Exposed to Pressure Pulses

DEFF Research Database (Denmark)

Andersen, Anders Peter; Mørch, Knud Aage

2012-01-01

at an extended water-solid interface by imposing a tensile stress pulse which easily causes cavitation. Next, a compressive pulse of duration ~1 ms and a peak intensity of a few bar is imposed prior to the tensile stress pulse. A dramatic increase of the tensile strength is observed immediately after......It is well known that pressurization for an extended period of time increases the tensile strength of water, but little information is available on the effect of pressure pulses of short duration. This is addressed in the present paper where we first measure the tensile strength of water...

Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

Energy Technology Data Exchange (ETDEWEB)

Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-30

The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000°C). Furthermore, these alloys have the potential to have prolonged survival under loss-of-coolant accident (LOCA) conditions compared to the more traditional cladding materials that are either Zr-based alloys or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility of mitigating welding-induced cracking via alloying and precise structure control of the weldments; in the frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing parameters. Also, a brief discussion of the irradiation at the High Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in detail focusing on the irradiation temperature. Limited fractography results are also presented and analyzed. The discussion highlights the limitations of the testing within a hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development and Analysis (LAMDA) laboratory and prepared for

Influence of some crosslinking agents on thermal and mechanical properties of electron beam irradiated polylactide

Energy Technology Data Exchange (ETDEWEB)

Rytlewski, Piotr, E-mail: prytlewski@ukw.edu.p [Department of Materials Engineering, Kazimierz Wielki University, ul. Chodkiewicza 30, 85-064 Bydgoszcz (Poland); Malinowski, RafaL [Institute for Engineering of Polymer Materials and Dyes, ul. M. SkLodowskiej-Curie 55, 87-100 Torun (Poland); Moraczewski, Krzysztof [Department of Materials Engineering, Kazimierz Wielki University, ul. Chodkiewicza 30, 85-064 Bydgoszcz (Poland); Zenkiewicz, Marian [Institute for Engineering of Polymer Materials and Dyes, ul. M. SkLodowskiej-Curie 55, 87-100 Torun (Poland)

2010-10-15

The aim of this article was to determine and compare the influence of trimethylopropane trimethacylate (TMPTA) and trially isocyanurate (TAIC) crosslinking agents on thermal and mechanical properties of electron beam irradiated polylactide (PLA). The blends were made of PLA mixed with 3 wt% of TMPTA (PLA/TMPTA), and PLA mixed with 3 wt% of TAIC (PLA/TAIC). Injection moulded samples were irradiated with the use of high energy (10 MeV) electron beam at various radiation doses to crosslinking PLA macromolecules. Thermal and mechanical properties were investigated by means of differential scanning calorimetry (DSC), dynamic mechanical analysis (DMA), tensile strength, and impact strength measurements. The samples were also characterized by Fourier transform infrared spectroscopy (FTIR). It was found that under the influence of electron irradiation PLA/TMPTA samples underwent degradation while PLA/TAIC samples became crosslinked. Tensile and impact strengths of PLA/TMPTA samples decreased with increasing radiation dose while an enhancement of these properties for PLA/TAIC samples was observed.

Influence of some crosslinking agents on thermal and mechanical properties of electron beam irradiated polylactide

International Nuclear Information System (INIS)

Rytlewski, Piotr; Malinowski, RafaL; Moraczewski, Krzysztof; Zenkiewicz, Marian

2010-01-01

The aim of this article was to determine and compare the influence of trimethylopropane trimethacylate (TMPTA) and trially isocyanurate (TAIC) crosslinking agents on thermal and mechanical properties of electron beam irradiated polylactide (PLA). The blends were made of PLA mixed with 3 wt% of TMPTA (PLA/TMPTA), and PLA mixed with 3 wt% of TAIC (PLA/TAIC). Injection moulded samples were irradiated with the use of high energy (10 MeV) electron beam at various radiation doses to crosslinking PLA macromolecules. Thermal and mechanical properties were investigated by means of differential scanning calorimetry (DSC), dynamic mechanical analysis (DMA), tensile strength, and impact strength measurements. The samples were also characterized by Fourier transform infrared spectroscopy (FTIR). It was found that under the influence of electron irradiation PLA/TMPTA samples underwent degradation while PLA/TAIC samples became crosslinked. Tensile and impact strengths of PLA/TMPTA samples decreased with increasing radiation dose while an enhancement of these properties for PLA/TAIC samples was observed.

High temperature tensile properties of 316 stainless steel implanted with helium

International Nuclear Information System (INIS)

Hasegawa, Akira; Yamamoto, Norikazu; Shiraishi, Haruki

1993-01-01

Helium embrittlement is one of the problems in structural materials for fusion reactors. Recently, martensitic steels have been developed which have a good resistance to high-temperature helium embrittlement, but the mechanism has not yet been clarified. In this paper, tensile behaviors of helium implanted austenitic stainless steels, which are sensitive to the helium embrittlement, were studied and compared with those of martensitic steels under the same experimental conditions, and the effect of microstructure on helium embrittlement was discussed. Helium was implanted by 300 appm at 573-623 K to miniature tensile speciments of 316 austenitic steels using a cyclotron accelerator. Solution annealed (316SA) and 20% cold worked (316CW) specimens were used. Post-implantation tensile tests were carried out at 573, 873 and 973 K. Yield stress at 573 K increased with the helium implantation in 316SA and 316CW, but the yield stress changes of 316SA at 873 and 973 K were different from that of 316CW. Black-dots were observed in the as-implanted specimen and bubbles were observed in the speciments tensile-tested at 873 and 973 K. Intergranular fracture was observed at only 973 K in both of the 316SA and 316CW specimens. Therefore, cold work did not suppress the high-temperature helium embrittlement under this experimental condition. The difference in the influence of helium on type 316 steel and 9Cr martensitic steels were discussed. Test temperature change of reduction in are showed clearly that helium embrittlement did not occur in 9Cr martensitic steels but occurred in 316 austenitic steels. Fine microstructures of 9Cr martensitic steels should suppress helium embrittlement at high temperatures. (author)

Mechanical properties of cladding tubes made from type 1.4970 SS after irradiation in a Rapsodie-bundle

International Nuclear Information System (INIS)

Schaefer, L.

1980-08-01

The mechanical properties of pin sections are tested in tensile and stress-rupture tests. The dependence of the tensile properties of the irradiation temperature, of the dosis of fast neutrons and of the deformation rate is described. The results are as expected except for a maximum of the yield strength at 400 0 C and a fluence of 2 x 10 22 (nsub(s)/cm 2 ). Stress-rupture tests have shown that the weakest part of the pin is at the hot end of the fuel column. There the stress-rupture strength is only 60% of the strength of an unirradiated tube, because of corrosion with fission products and other influences. Taking into account the loss of cross section due to corrosion, the stress-rupture strength of pin sections agree with that of specimens from material irradiation experiments. The ductility is above 0.2% in the stress rupture test and above 0.5% in the tensile test. (orig.) [de

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