Tensile properties of neutron irradiated 316Ti and 15-15Ti steels

International Nuclear Information System (INIS)

Fissolo, A.; Levy, V.; Seran, J.L.; Maillard, A.; Royer, J.; Rabouille, O.

1992-01-01

This paper deals with the tensile behavior of CW316Ti and CW15-15Ti Phenix fuel pin cladding. The tensile tests were conducted on defueled tubes irradiated up to 115 dpa 3 in the 400-640 deg C temperature range. Test temperature corresponds essentially to irradiation temperature. The results emphasize that although irradiation induces a reduction of ductility, failure always occurs with significant plastic deformation even for the most irradiated clads. (author). 15 refs., 12 figs., 1 tab

Tensile properties in zircaloy-II after 590 MeV proton irradiation

Energy Technology Data Exchange (ETDEWEB)

Dai, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Victoria, M. [Ecole Polytechnique Federale, Lausanne (Switzerland)

1997-09-01

In order to investigate radiation potential damage effects on the SINQ Zircaloy-rod target, four Zircaloy-II tensile specimens were irradiated at the PIREX facility in 1995 to a proton fluence about 3x10{sup 20} p/cm{sup 2}, which produced a radiation damage of about 1.35 displacements per atom (dpa). Tensile test results show that, although there is some reduction in tensile elongation, substantial ductility still exists after such irradiation dose which corresponds to the peak value obtained in the SINQ target for 23 days operation at 1 mA. (author) 1 fig., 2 refs.

Assessment of models predicting irradiation effects on tensile properties of reactor pressure vessel steels

International Nuclear Information System (INIS)

Pineau, L.; Landron, C.

2015-01-01

In this paper, an analysis of tensile data acquired as part of the French Reactor Vessel Surveillance Program (RVSP) is produced. This program contains amongst other mechanical tests, tensile tests at 20 and 300 C degrees on non irradiated base metals and at 300 C degrees only on irradiated materials. It shows that irradiation leads to an increase in the yield strength and a decrease in the strain hardening. The exploitation of tensile results has permitted to express a relationship between yield strength increase measured and fluence value, as well as between strain hardening decrease and yield strength evolution. The use of these relations in the aim at predicting evolution of tensile properties with irradiation has then permitted to propose a methodology to model entire stress-strain curves of irradiated base metal only based on the non irradiated stress-strain curve. These predictions were successfully compared with an experimental standard case. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Tsai, H.C.; Nowicki, L.J. [Argonne National Lab., IL (United States)] [and others

1997-08-01

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

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.

Effect of cold work on tensile behavior of irradiated type 316 stainless steel

International Nuclear Information System (INIS)

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

1986-01-01

Tensile specimens were irradiated in ORR at 250, 290, 450, and 500 0 C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290 0 C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450 0 C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500 0 C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500 0 C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes

Tensile tests and metallography of brazed AISI 316L specimens after irradiation

International Nuclear Information System (INIS)

Groot, P.; Franconi, E.

1994-01-01

Stainless steel type 316L tensile specimens were vacuum brazed with three kinds of alloys: BNi-5, BNi-6, and BNi-7. The specimens were irradiated up to 0.7 dpa at 353 K in the High Flux Reactor at JRC Petten, the Netherlands. Tensile tests were performed at a constant displacement rate of 10 -3 s -1 at room temperature in the ECN hot cell facility. BNi-5 brazed specimens showed ductile behaviour. Necking and fractures were localized in the plate material. BNi-6 and BNi-7 brazed specimens failed brittle in the brazed zone. This was preceded by uniform deformation of the plate material. Tensile test results of irradiated specimens showed higher stresses due to radiation hardening and a reduction of the elongation of the plate material compared to the reference. SEM examination of the irradiated BNi-6 and BNi-7 fracture surfaces showed nonmetallic phases. These phases were not found in the reference specimens. ((orig.))

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.

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

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 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

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.

Tensile properties of irradiated TZM and tungsten

International Nuclear Information System (INIS)

Steichen, J.M.

1975-04-01

The effect of neutron irradiation on the elevated temperature tensile properties of TZM and tungsten has been experimentally determined. Specimens were irradiated at a temperature of approximately 720 0 F to fluences of 0.4 and 0.9 x 10 22 n/cm 2 (E greater than 0.1 MeV). Test parameters for both control and irradiated specimens included strain rates from 3 x 10 -4 to 1 s -1 and temperatures from 72 to 1700 0 F. The results of these tests were correlated with a rate-temperature parameter (T ln A/epsilon) to provide a concise description of material behavior over the range of deformation conditions of this study. The yield strength of the subject materials was significantly increased by decreasing temperature, increasing strain rate, and increasing fluence. Ductility was significantly reduced at any temperature or strain rate by increasing fluence. Cleavage fractures occurred in both unirradiated and irradiated specimens when the yield strength was elevated to the effective cleavage stress by temperature and/or strain rate. Neutron irradiation for the conditions of this study increased the ductile-to-brittle transition temperature of tungsten by approximately 300 0 F and TZM by approximately 420 0 F. (U.S.)

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.)

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...

Post-irradiation diarrhea

International Nuclear Information System (INIS)

Meerwaldt, J.H.

1984-01-01

In radiotherapy of pelvic cancers, the X-ray dose to be delivered to the tumour is limited by the tolerance of healthy surrounding tissue. In recent years, a number of serious complications of irradiation of pelvic organs were encountered. Modern radiotherapy necessitates the acceptance of a calculated risk of complications in order to achieve a better cure rate. To calculate these risks, one has to know the radiation dose-effect relationship of normal tissues. Of the normal tissues most at risk when treating pelvic tumours only the bowel is studied. In the literature regarding post-irradiation bowel complications, severe and mild complications are often mixed. In the present investigation the author concentrated on the group of patients with relatively mild symptoms. He studied the incidence and course of post-irradiation diarrhea in 196 patients treated for carcinoma of the uterine cervix or endometrium. The aims of the present study were: 1) to determine the incidence, course and prognostic significance of post-irradiation diarrhea; 2) to assess the influence of radiotherapy factors; 3) to study the relation of bile acid metabolism to post-irradiation diarrhea; 4) to investigate whether local factors (reservoir function) were primarily responsible. (Auth.)

Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

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.

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

Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

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

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

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

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

Post irradiation test report of irradiated DUPIC simulated fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

2001-12-01

The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

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)

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.

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.

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

2000-12-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

2001-09-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

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).

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...

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)

Post irradiation examination technology exchange

International Nuclear Information System (INIS)

Sozawa, Shizuo; Ito, Masayasu; Taguchi, Taketoshi; Nakagawa, Tetsuya; Lee, Hyung-Kwon

2012-01-01

Under the KAERI and JAEA agreement, in a part of the program 18 (Post Irradiation Examination (PIE) and Evaluation Technique of Irradiated Materials), an eddy current test was proposed as a round robin test, and it has been being progressed in both organizations in order to enhance the post irradiation examination technology. Up to now, several data are obtained by both PIE facilities. In this paper, the round robin test program is shown, and also shown obtained data with discussion from applicability as a nondestructive test in the hot cell. (author)

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

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.)

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)

Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

International Nuclear Information System (INIS)

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

1985-01-01

Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

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

AGR-1 Post Irradiation Examination Final Report

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-08-01

The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

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

Selective Laser Melting Produced Ti-6Al-4V: Post-Process Heat Treatments to Achieve Superior Tensile Properties.

Science.gov (United States)

Ter Haar, Gerrit M; Becker, Thorsten H

2018-01-17

Current post-process heat treatments applied to selective laser melting produced Ti-6Al-4V do not achieve the same microstructure and therefore superior tensile behaviour of thermomechanical processed wrought Ti-6Al-4V. Due to the growing demand for selective laser melting produced parts in industry, research and development towards improved mechanical properties is ongoing. This study is aimed at developing post-process annealing strategies to improve tensile behaviour of selective laser melting produced Ti-6Al-4V parts. Optical and electron microscopy was used to study α grain morphology as a function of annealing temperature, hold time and cooling rate. Quasi-static uniaxial tensile tests were used to measure tensile behaviour of different annealed parts. It was found that elongated α'/α grains can be fragmented into equiaxial grains through applying a high temperature annealing strategy. It is shown that bi-modal microstructures achieve a superior tensile ductility to current heat treated selective laser melting produced Ti-6Al-4V samples.

Effects of post-irradiation annealing on the transformation behavior of Ti-Ni alloys

International Nuclear Information System (INIS)

Kimura, A.; Tsuruga, H.; Morimura, T.; Misawa, T.; Miyazaki, S.

1993-01-01

Recovery processes of martensitic transformation of neutron irradiated Ti-50.0, 50.5 and 51.0 at.%Ni alloys during post-irradiation annealing were investigated by means of differential scanning calorimetry (DSC), tensile tests and transmission electron microscope (TEM) observations. Neutron irradiation up to a fluence of 1.2x10 24 n/cm 2 at 333 K suppressed the martensitic transformation as well as the stress-induced martensitic transformation of these alloys above 150 K. The TEM observations revealed that the disordered zones containing small defect clusters in high density were formed in the neutron irradiated Ti-Ni alloys. The DSC measurements also showed that the post-irradiation annealing caused recovery of the transformation of which the progress depended on the annealing temperature and period. A significant retardation of the recovery was recognized in the Ti-51.0 at.%Ni alloy in comparison with the Ti-50.0 at.%Ni alloy. From the shifts in the transformation temperature upon isothermal annealing at various annealing temperatures, the activation energies of the recovery process of the transformation in the neutron irradiated Ti-50.0 and 51.0 at.%Ni alloys were evaluated by a cross-cut method to be 1.2 eV and 1.5 eV, respectively. The recovery of the transformation was ascribed to the re-ordering resulting from decomposition of vacancy clusters, and those obtained values of the activation energy were considered to be the sum of the migration energy of vacancy and the binding energy of vacancy-vacancy cluster. The retardation of the recovery in the Ti-51.0 at%Ni alloy was interpreted in terms of large binding energy in this alloy due to the off-stoichiometry. (author)

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])

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

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

Post-irradiation examination of fuel elements of Tarapur Atomic Power Station (Report-I)

International Nuclear Information System (INIS)

Bahl, J.K.; Sah, D.N.; Chatterjee, S.; Sivaramkrishnan, K.S.

1979-01-01

Detailed post-irradiation examination of three initial load fuel elements of the Tarapur Atomic Power Station (TAPS) has been carried out. The causes of the element failures have been analysed. It was observed that almost 90% of the length of the elements exoerienced nodular corrosion. It has been estimated that nodular corrosion would seriously affect the wall thickness and surface temperature of higher rated elements. Lunar shaped fret marks have also been observed at some spacer grid locations in the elements. The depth of the largest fret mark was measured to be 16.9% clad wall thickness. Detailed metallographic examination of the clad and fuel in the three elements has been done. The temperatures at different structural regions of the fuel cross-sections have been estimated. The change in fuel density during irradiation has been evaluated by comparing the irradiated fuel diameter with the mean pellet design diameter. The performance of the end plug welds and spacer grid sites in the elements has been assessed. The burnup distribution along the length of the elements has been evaluated by gamma scanning. The redistribution of fission products in the fuel has been examined by gamma scanning and beta-gamma autoradiography. Mechanical properties of the irradiated cladding have been examined by ring tensile testing. (auth.)

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

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

Selective Laser Melting Produced Ti-6Al-4V: Post-Process Heat Treatments to Achieve Superior Tensile Properties

Directory of Open Access Journals (Sweden)

Gerrit M. Ter Haar

2018-01-01

Full Text Available Current post-process heat treatments applied to selective laser melting produced Ti-6Al-4V do not achieve the same microstructure and therefore superior tensile behaviour of thermomechanical processed wrought Ti-6Al-4V. Due to the growing demand for selective laser melting produced parts in industry, research and development towards improved mechanical properties is ongoing. This study is aimed at developing post-process annealing strategies to improve tensile behaviour of selective laser melting produced Ti-6Al-4V parts. Optical and electron microscopy was used to study α grain morphology as a function of annealing temperature, hold time and cooling rate. Quasi-static uniaxial tensile tests were used to measure tensile behaviour of different annealed parts. It was found that elongated α’/α grains can be fragmented into equiaxial grains through applying a high temperature annealing strategy. It is shown that bi-modal microstructures achieve a superior tensile ductility to current heat treated selective laser melting produced Ti-6Al-4V samples.

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,

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.

Post irradiation examinations on HTTR materials

International Nuclear Information System (INIS)

Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

1995-01-01

The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

Post-irradiation stability of polyvinyl chloride at sterilizing doses

International Nuclear Information System (INIS)

Naimian, F.; Katbab, A.A.; Nazokdast, H.

1994-01-01

Post-irradiation stability of plasticized PVC irradiated by 60 Co gamma ray at sterilizing doses has been studied. Effects of irradiation upon chemical structure, mechanical properties and rheological behaviour of samples contained different amounts of Di(2-ethylhexyl)phthalate as plasticizer have been investigated. Formation of conjugated double bonds, carbonyl and hydroxyl groups have been followed by UV and FTIR spectrometers up to 6 months after irradiation. FTIR spectra of irradiated samples showed no significant changes in carbonyl and hydroxyl groups even 6 months after irradiation. However, changes in UV-visible spectra was observed for the irradiated samples up to 6 months post-irradiation. This has been attributed to the formation of polyenes which leads to the discoloration of this polymer. Despite a certain degree of discoloration, it appears that the mechanical properties of PVC are not affected by irradiation at sterilizing doses. No change in the melt viscosity of the irradiated PVC samples with post-irradiation was observed, which is inconsistent with the IR results. (author)

New Therapeutic Possibilities of the Post-Irradiation Haemorrhagic Syndrome

Energy Technology Data Exchange (ETDEWEB)

Pospisil, J.; Dienstbier, Z. [Institute of Biophysics and Nuclear Medicine, Faculty of General Medicine, Charles University, Prague, Czechoslovak Socialist Republic (Czech Republic); Skala, E. [Central Military Hospital, Prague-Stresovice, Czechoslovak Socialist Republic (Czech Republic)

1969-10-15

Haemorrhagic diathesis is one of the dominant symptoms of acute post-irradiation lesion. Haemorrhagic syndrome is caused by the disturbance of haemocoagulation during simultaneous lesion of the vascular system. In our study we have tried to affect the post-irradiation haemocoagulation disturbance. Epsilon- amino-caproic acid (EACA) administered between the 8{sup th} and the 18{sup th} day (0.4 g/kg per day) to whole- body irradiated dogs (600 R) partially regulated the post-irradiation disturbance of haemocoagulation. The favourable effect of EACA was verified by in vitro experiments in which the blood of irradiated dogs was used. A repeated administration of EACA in the dose of 0.4 g/kg per day to whole-body irradiated rats (600 R) did not substantially affect the post-irradiation changes in the number of white blood elements; however, its administration to healthy animals caused lymphocytosis. In whole-body irradiated dogs (600 R) we have found lower levels of EACA in the blood up to the 8 day following irradiation as compared with healthy dogs after oral application of EACA. The whole-body irradiation of mice did not increase the acute toxicity of EACA. The daily administration of 0.4 g EACA/kg to whole-body irradiated mice (600 and 700 R) did not change the mortality induced by irradiation. The authors consider EACA to be a suitable compound for a complex therapy of radiation sickness. The administration of para-amino-methyl-benzoic acid (PAMBA), in spite of a certain improvement of postirradiation haemocoagulation disturbance, is less efficient. Our recent experiments with ellagic acid which significantly affects the post-traumatic haemorrhage in whole-body irradiated rats seem to be very promising. (author)

Chemical changes after irradiation and post-irradiation storage in tilapia and Spanish mackerel

International Nuclear Information System (INIS)

Al-Kahtani, H.A.; Abu-Tarboush, H.M.; Bajaber, A.S.; Atia, M.; Abou-Arab, A.A.; El-Mojaddidi, M.A.

1996-01-01

Influence of gamma irradiation (1.5-10 kGy) and post-irradiation storage up to 20 days at 2 +/- 2 degrees C on some chemical criteria of tilapia and spanish mackerel were studied. Total volatile basic nitrogen formation was lower in irradiated fish than in the unirradiated. Irradiation also caused a larger increase in thiobarbituric acid values which continued gradually during storage. Some fatty acids decreased by irradiation treatments at all doses. Thiamin loss was more severe at higher doses (greater than or equal to 4.5 kGy), whereas riboflavin was not affected. Alpha and gamma tocopherols of tilapia and alpha, beta, gamma, and delta tocopherols, in Spanish mackerel, decreased with increased dose and continued to decrease during 20-day post-irradiation storage

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.

AGC-2 Specimen Post Irradiation Data Package Report

Energy Technology Data Exchange (ETDEWEB)

Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-08-01

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

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

Waiting Time for Coronal Preparation and the Influence of Different Cements on Tensile Strength of Metal Posts

Directory of Open Access Journals (Sweden)

Ilione Kruschewsky Costa Sousa Oliveira

2012-01-01

Full Text Available This study aimed to assess the effect of post-cementation waiting time for core preparation of cemented cast posts and cores had on retention in the root canal, using two different luting materials. Sixty extracted human canines were sectioned 16 mm from the root apex. After cast nickel-chromium metal posts and cores were fabricated and luted with zinc phosphate (ZP cement or resin cement (RC, the specimens were divided into 3 groups (n = 10 according to the waiting time for core preparation: no preparation (control, 15 minutes, or 1 week after the core cementation. At the appropriate time, the specimens were subjected to a tensile load test (0.5 mm/min until failure. Two-way ANOVA (time versus cement and the Tukey tests (P < 0.05 showed significantly higher (P < 0.05 tensile strength values for the ZP cement groups than for the RC groups. Core preparation and post-cementation waiting time for core recontouring did not influence the retention strength. ZP was the best material for intraradicular metal post cementation.

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

2016-11-01

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

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)

Role of post irradiation growth delay in chemical radioprotection by caffeine

International Nuclear Information System (INIS)

Gangabhagirathi, R.; Rao, B.S.; Bhat, N.N.

2004-01-01

Post irradiation treatment with caffeine enhanced the survival of wild type diploid yeast strain, Saccharomyces cerevisiae X2180. The presence of caffeine during gamma irradiation also affected a similar enhancement in survival. These observations suggest that caffeine imparted significant protection against radiation. Effectiveness of caffeine, even when present only during the post irradiation period, suggests that it modulates the post irradiation recovery process in yeast cells. (author)

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

16-rod-bundle: Irradiation in the MZFR and post-irradiation examinations

International Nuclear Information System (INIS)

Manzel, R.

1979-04-01

In the course of the irradiation of a 16-rod prototype bundle, the basis has been established for the irradiation of experimental fuel assemblies containing full-length PWR fuel rods in standard positions of the MZFR. The prototype bundle was discharged after an irradiation time of 284 full power days and a burnup of 11400 MWd/tU. The overall performance of the prototype bundle was highly satisfactory. Detailed post-irradiation examinations confirmed the good conditions of bundle structures and fuel rods. (orig.) [de

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.)

Effect of medium and post-irradiation storage on rooting of irradiated onions

International Nuclear Information System (INIS)

Singh, Rita

2000-01-01

Rooting test for detection of irradiation in onion bulbs was studied. Onions were exposed to different dose levels of 30, 60, 90, 120 and 150 Gy. The effects of irradiation dose, cultivar difference, rooting medium and post-irradiation storage on the rooting were investigated. The number and the length of the roots formed in onions were found to decrease on irradiation. The effect was more at higher doses. The effect of irradiation on rooting was also evident after 120 days of storage. (author)

The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

International Nuclear Information System (INIS)

Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

1995-01-01

This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

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

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.

Evaluation of post-operative prophylactic irradiation for carcinoma of the esophagus

International Nuclear Information System (INIS)

Mafune, Ken-ichi; Tanaka, Yoichi; Fujita, Kichishiro; Sakura, Mizuyoshi

1987-01-01

Of 147 patients with carcinoma of the esophagus resected at Saitama Cancer Center Hospital for 10 years, 98 cases were studied to evaluate post-operative prophylactic irradiation. The total dose of irradiation was up to 4,000 ∼ 5,000 rads of Linac X-ray and the irradiated field was T-shaped covering the upper mediastinal and bilateral cervical regions. The prognosis of the post-operative irradiated group (56 cases) was significantly better than that of the control group (42 cases) (p < 0.01). This study resulted in a five-year survival rate of 34.2 percent for patients in the post-operative irradiated group, compared to 16.7 percent for those in the control group. Further detailed comparative studies revealed similar results. Cancer recurrence occurred at the irradiated fields in 8 cases (14.3 %), though in 15 cases (35.7 %) of the control group. This suggested the local suppressive effect of the post-operative irradiation to the cancer recurrence. (author)

Post-irradiation characterization of PH13-8Mo martensitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Jong, M.; Schmalz, F.; Rensman, J.W. [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Luzginova, N.V., E-mail: luzginova@nrg.eu [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Wouters, O.; Hegeman, J.B.J.; Laan, J.G. van der [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

2011-10-01

The irradiation response of PH13-8Mo stainless steel was measured up to 2.5 dpa at 200 and 300 deg. C irradiation temperatures. The PH13-8Mo, a martensitic precipitation-hardened steel, was produced by Hot Isostatic Pressing at 1030 deg. C. The fatigue tests (high cycle fatigue and fatigue crack propagation) showed a test temperature dependency but no irradiation effects. Tensile tests showed irradiation hardening (yield stress increase) of approximately 37% for 200 deg. C irradiated material tested at 60 deg. C and approximately 32% for 300 deg. C irradiated material tested at 60 deg. C. This contradicts the shift in reference temperature (T{sub 0}) measured in toughness tests (Master Curve approach), where the {Delta}T{sub 0} for 300 deg. C irradiated is approximately 170 deg. C and the {Delta}T{sub 0} for the 200 deg. C irradiated is approximately 160 deg. C. This means that the irradiation hardening of PH13-8Mo steel is not suitable to predict the shift in the reference temperature for the Master Curve approach.

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

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

Mechanical Properties of Post Irradiation Primary Cooling Piping of Bandung Research Reactor

International Nuclear Information System (INIS)

Histori; Renaningsih S; Sri Nitiswati; Ari Triyadi

2003-01-01

Testing on primary coolant piping of research reactor Bandung have been done. Primary coolant piping were made from Al 6061-T6. The goal of this activity is to investigate the mechanical properties changes caused by aging process after 33 years in irradiated. Type of testing i.e visual examination, thickness measurement, tensile and hardness test were done. The test data shown that there was a deposit at the inside surface of pipe, thickness decreased about 0.2 mm, tensile strength is 293 MPa, yield strength is 262 MPa, while the hardness is about 83 HRE (mean value). The test data than compared with ASTM standard. As the conclusion tensile and yield strength of pipe still fulfill the ASTM requirements, except the hardness is unsignificantly less/decreased. (author)

Preliminary test results for post irradiation examination on the HTTR fuel

International Nuclear Information System (INIS)

Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

2007-01-01

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

Hair transplantation for the the treatment of post-irradiation alopecia

International Nuclear Information System (INIS)

Kolasinski, J.; Kolenda, M.; Skowronek, J.

2002-01-01

Treatment of head and neck tumours and of leukaemia often necessitates radiotherapy. However; permanent alopecia in the scalp exposed to irradiation is a common problem. One of the effective methods of treatment of post-irradiation alopecia is hair transplantation. Over a period of 18 years 42 patients were treated at the Hair Clinic Poznan for post-irradiation alopecia. Due to the presence of numerous lesions in the donor and recipient scalp areas many modifications were introduced into alopecia correction. The treatment assured good cosmetic effects, free of the risk of complications. Scalps from occipital areas do not go bald when transferred to scalp areas affected by balding. On the contrary - they retain original properties, thus resulting in hair re-growth. Hair follicle transplantation is usually applied for the correction of androgenic alopecia in men and women although it may also be applied in post-trauma and post-irradiation alopecia treatment. Hair regrowth in radiotherapy patients occurs later than in androgenic alopecia patients. This phenomenon is caused by blood supply deficits in the recipient area. Autogenic hair follicle transplantation is a treatment of choice in the correction of post-irradiation alopecia, while the good cosmetic effects considerably improve the patients' quality of life. (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.

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

Influence of core-finishing intervals on tensile strength of cast posts-and-cores luted with zinc phosphate cement

Directory of Open Access Journals (Sweden)

Michele Andrea Lopes Iglesias

2012-08-01

Full Text Available The core finishing of cast posts-and-cores after luting is routine in dental practice. However, the effects of the vibrations produced by the rotary cutting instruments over the luting cements are not well-documented. This study evaluated the influence of the time intervals that elapsed between the cementation and the core-finishing procedures on the tensile strength of cast posts-and-cores luted with zinc phosphate cement. Forty-eight bovine incisor roots were selected, endodontically treated, and divided into four groups (n = 12: GA, control (without finishing; GB, GC, and GD, subjected to finishing at 20 minutes, 60 minutes, and 24 hours after cementation, respectively. Root canals were molded, and the resin patterns were cast in copper-aluminum alloy. Cast posts-and-cores were luted with zinc phosphate cement, and the core-finishing procedures were applied according to the groups. The tensile tests were performed at a crosshead speed of 0.5 mm/min for all groups, 24 hours after the core-finishing procedures. The data were subjected to one-way analysis of variance (ANOVA and Tukey's test (α = 0.05. No significant differences were observed in the tensile strengths between the control and experimental groups, regardless of the time interval that elapsed between the luting and finishing steps. Within the limitations of the present study, it was demonstrated that the core-finishing procedures and time intervals that elapsed after luting did not appear to affect the retention of cast posts-and-cores when zinc phosphate cement was used.

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)

Post irradiation examination of garter springs from Indian PHWRs

International Nuclear Information System (INIS)

Dubey, J.S.; Shah, Priti Kotak; Mishra, Prerna; Singh, H.N.; Alur, V.D.; Kumar, Ashwini; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

Irradiated Zr-2.5Nb-0.5Cu garter springs, belonging to Indian Pressurised Heavy Water Reactors, which had experienced 8 to 10 Effective Full Power Years of operation were subjected to visual, dimensional, chemical, metallographic examination and relevant mechanical tests. Methodology of the tests conducted and results are presented. The digital photographs were used to measure the inner and outer circumferences by image processing. The hydrogen (H) content in the spring coils were measured using Differential Scanning Calorimetry (DSC). In the stretch test, all the irradiated GSs were found to require an additional load, as compared to unirradiated GS, to produce a given amount of residual extension which indicated that the irradiated GSs had undergone significant irradiation hardening. The crush test results showed that the minimum load required to crush the coil or cause a sudden sideways shift in the grips was higher than 400 N/coil, much higher than the design load. The test results indicated that the irradiated GS, after 10 EFPY of operation, have adequate strength and ductility to continue to meet the design intent. Mechanical tests were carried out on irradiated girdle wires taken out of the loose fit garter springs (GS) from (NAPS-1, ∼ 8.5 EFPY) and tight fit garter spring from KAPS-2 (∼ 8.0 EFPY) PHWRs. Tensile tests on the irradiated girdle wires, showed irradiation hardening in the material and reduction in ductility. The irradiated girdle wires have around 4 to 5% residual ductility level against the 15% ductility of unirradiated wire. The fracture surfaces of the irradiated as well as the un-irradiated girdle wires were observed in SEM. (author)

UV irradiation improves the bond strength of resin cement to fiber posts.

Science.gov (United States)

Zhong, Bo; Zhang, Yong; Zhou, Jianfeng; Chen, Li; Li, Deli; Tan, Jianguo

2011-01-01

The purpose is to evaluate the effect of UV irradiation on the bond strength between epoxy-based glass fiber posts and resin cement. Twelve epoxy-based glass fiber posts were randomly divided into three groups. Group 1 (Cont.): No surface treatment. Group 2 (Low-UV): UV irradiation was conducted from a distance of 10 cm for 10 min. Group 3 (High-UV): UV irradiation was conducted from a distance of 1 cm for 3 min. A resin cement (CLEARFIL SA LUTING) was used for the post cementation to form resin slabs which contained fiber posts in the center. Microtensile bond strengths were tested and the mean bond strengths (MPa) were 18.81 for Cont. group, 23.65 for Low-UV group, 34.75 for High-UV group. UV irradiation had a significant effect on the bond strength (pUV irradiation demonstrates its capability to improve the bond strength between epoxy-based glass fiber posts and resin cement.

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

Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

International Nuclear Information System (INIS)

Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

2010-11-01

The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (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

1979-01-01

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)

Post irradiation examination on test fuel pins for PWR

International Nuclear Information System (INIS)

Fogaca Filho, N.; Ambrozio Filho, F.

1981-01-01

Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

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)

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

Irradiation and Post-Irradiation Storage of Chicken: Effects on Fat and Proteins

International Nuclear Information System (INIS)

Abou-Tarboush, H.M.; Al-Kahtani, H.A.; Abou-Arab, A.A.; Atia, M.; Bajaber, A.S.; Ahmed, M.A.; El-Mojaddidi, M.A.

1997-01-01

Chicken were subjected to gamma irradiation doses of 2.5, 5.0, 7.5 and 10.0 KGy and post-irradiation storage of 21 days at 4±2º. The effects on fat and protein of chicken were studied. Rate of formation of total volatile basic-nitrogen was less in irradiated samples particularly in samples treated with 5.0KGy during the entire storage. Fatty acid profiles of chicken lipids were not significantly (P≤ 0.05) affected by irradiation especially at doses of 5.0 KGy. However, irradiation caused a large increase in thiobarbituric acid (TBA) values which continued gradually during storage. Changes in amino acids were minimal. Irradiated and unirradiated samples showed the appearance of protein subunits with molecular weights in the range of 10.0 to 88.0 and 10.0 to 67.0 KD, respectively. No changes were observed in the sarcoplasmic protein but the intensity of bands in all irradiated samples decreased after 21 days of storage

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

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

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.

Technical review on irradiation tests and post-irradiation examinations in JMTR

International Nuclear Information System (INIS)

2017-07-01

The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

Post-irradiation replication and repair in UV-irradiated cells of Proteus mirabilis depends on protein synthesis and a functioning rec+ gene

International Nuclear Information System (INIS)

Hofemeister, J.

1977-01-01

The amount of and the molecular weight of newly synthesized DNA (piDNA) as well as its repair after UV irradiation in excision-proficient strains of P.mirabilis and E.coli K12 have been compared. A fraction of post-replication repair (PRR) in P.mirabilis is found to be dependent on de novo protein synthesis after UV irradiation. Pre-irradiation by UV and pre-treatment with nalidixic acid increase the efficiency of post-irradiation replication and PRR even in the presence of chloramphenicol. An inducible repair function in P.mirabilis is supposed to stimulate post-irradiation replication and repair. (author)

Effect of post weld heat treatment on tensile properties and microstructure characteristics of friction stir welded armour grade AA7075-T651 aluminium alloy

OpenAIRE

Sivaraj, P.; Kanagarajan, D.; Balasubramanian, V.

2014-01-01

This paper reports the effects of post weld heat treatments, namely artificial ageing and solution treatment followed by artificial ageing, on microstructure and mechanical properties of 12 mm thick friction stir welded joints of precipitation hardenable high strength armour grade AA7075-T651 aluminium alloy. The tensile properties, such as yield strength, tensile strength, elongation and notch tensile strength, are evaluated and correlated with the microhardness and microstructural features....

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.

Cytogenetics of Post-Irradiation Mouse Leukaemia

Energy Technology Data Exchange (ETDEWEB)

Wald, N.; Pan, S.; Upton, A.; Brown, R. [Graduate School of Public Health, University of Pittsburgh, PA (United States); Oak Ridge National Laboratory, Oak Ridge, TN (United States)

1969-11-15

The interrelationship between radiation, cytogenetic abnormalities, and viruses in leukaemogenesis has been studied in the RF/Un mouse which develops a high incidence of granulocytic leukaemia on radiation exposure. A virus-like agent has been demonstrated in such leukaemic animals and the disease has been transmitted by passage of apparently acellular materials from irradiated primary animals to normal recipients. Pilot cytogenetic studies revealed consistent abnormal chromosome markers and modal shifts in both irradiated leukaemic animals and in non-irradiated animals developing leukaemia after passage injection. To define better the relationship between consistent bone-marrow chromosome aberrations and postirradiation primary and passaged leukaemia, 100 RF/Un mice were studied which were irradiated with 300 R of 250-kVp X-rays at 100 weeks of age and subsequently developed leukaemia. Eighty-seven had granulocytic leukaemia and in 72 of these, bone-marrow cytogenetic abnormalities were found. The distribution of-numerical and structural chromosome aberrations in 3225 cells studied are reviewed in derail. The correlation of specific aberrations to clinical and histopathologic findings has been attempted: Sequential passages of apparently cell-free material from the post-irradiation leukaemic mice into unirradiated RE/Un recipients and subsequent passages from leukaemic recipients were performed to observe the evolution of any initial chromosome markers and shifts in modal chromosome number in the passage generations. Two-hundred-thirty-six mice were inoculated with the material obtained either from primary post-irradiation leukaemic mice or from serially-passaged leukaemia cases. In the most extensive passaged line, 22 transfer generations containing 129 leukaemic mice were examined by clinical, histopathologic, -haematologic and cytogenetic procedures. Evolution of abnormal chromosome modes from 41 in the early passages to 39 chromosomes consistently after the 4

Effects of material property changes on irradiation assisted stress corrosion cracking

Energy Technology Data Exchange (ETDEWEB)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2002-09-01

Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10{sup 26}n/m{sup 2} (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10{sup 24}n/m{sup 2} (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

Effects of material property changes on irradiation assisted stress corrosion cracking

International Nuclear Information System (INIS)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko

2002-01-01

Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10 26 n/m 2 (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10 24 n/m 2 (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

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

Post-irradiation data analysis for NRC/PNL Halden assembly IFA-431

International Nuclear Information System (INIS)

Nealley, C.; Lanning, D.D.; Cunningham, M.E.; Hann, C.R.

1979-10-01

Results are presented for the post irradiation examination performed on IFA-431, which was a 6-rod test fuel assembly irradiated in Halden Reactor, Norway, under sponsorship of the Nuclear Regulatory Commission. The irradiation conditions included: peak powers of 33 kW/m; coolant pressure and temperature of 3.3 MPa and 240 0 C, respectively; and peak burnup of 4300 MWd/MTM. IFA-431 included instrumented rods of basic boiling water reactor design, with variations in fill gas composition, gap size, and UO 2 fuel type. The irradiation was designed to measure the effect of these variations upon fuel rod thermal and mechanical performance. The post irradiation examination assessed the permanent changes to the rods, including induced radioactivity, cladding deformation, fission gas release, and fuel densification

Irradiation as an alternative post harvest treatment

Energy Technology Data Exchange (ETDEWEB)

Satin, M. [Agricultural Industries and Post-harvest Management Service, FAO, Rome (Italy); Loaharanu, P. [Head, Food Preservation Section, Joint FAO/ IAEA Division of Nuclear Techniques in Food and Agriculture, Wagramerstr. 5, A-1400, Vienna (Austria)

1997-12-31

This current world population has significantly added to the pressures placed upon our finite resources and our resulting ability to feed ourselves. In order to cope with current and future demands, the two established lines of action, that is, reduced population growth and expansion of agricultural production, must be supplemented with the parallel activity of reducing food losses during and after harvest. For developing countries in particular, enormous post-harvest losses result from spillage, contamination, pests and physiological deterioration during storage. Studies in these countries indicate that post-harvest losses are enormous and amount to tens of millions of tons per year valued at billions of dollars. Programs to reduce post-harvest losses, if applied properly, can result in realistic yield increases between 10 and 30%, which can be directly converted into increased consumption for humans. Post-harvest losses vary greatly and are a function of the crop variety, pest combinations in the environment, climate, the system of harvesting, storage, handling, marketing, and even the social and cultural environment. Pests are among the most criticals of these factors. Because of the disastrous potential consequences of such pests, quarantine regulations prohibit the entrance of plants or products which might hide the unwanted pest from countries where it is known to exist. Quarantine treatments are can be chemical, physical or ionizing radiation treatment. Numerous investigations on the use of ionizing radiation for the disinfestation of fresh plant materials indicate that rather low dosages will control fruit-fly problems, thus making it well suited for quarantine treatment. The effectiveness of the irradiation as a broad spectrum quarantine treatment of fresh fruits and vegetables was recognized by the several plant protection organizations around the world. Currently, some 40 countries have approved one or more irradiated food items or groups of food

Irradiation as an alternative post harvest treatment

International Nuclear Information System (INIS)

Satin, M.; Loaharanu, P.

1997-01-01

This current world population has significantly added to the pressures placed upon our finite resources and our resulting ability to feed ourselves. In order to cope with current and future demands, the two established lines of action, that is, reduced population growth and expansion of agricultural production, must be supplemented with the parallel activity of reducing food losses during and after harvest. For developing countries in particular, enormous post-harvest losses result from spillage, contamination, pests and physiological deterioration during storage. Studies in these countries indicate that post-harvest losses are enormous and amount to tens of millions of tons per year valued at billions of dollars. Programs to reduce post-harvest losses, if applied properly, can result in realistic yield increases between 10 and 30%, which can be directly converted into increased consumption for humans. Post-harvest losses vary greatly and are a function of the crop variety, pest combinations in the environment, climate, the system of harvesting, storage, handling, marketing, and even the social and cultural environment. Pests are among the most criticals of these factors. Because of the disastrous potential consequences of such pests, quarantine regulations prohibit the entrance of plants or products which might hide the unwanted pest from countries where it is known to exist. Quarantine treatments are can be chemical, physical or ionizing radiation treatment. Numerous investigations on the use of ionizing radiation for the disinfestation of fresh plant materials indicate that rather low dosages will control fruit-fly problems, thus making it well suited for quarantine treatment. The effectiveness of the irradiation as a broad spectrum quarantine treatment of fresh fruits and vegetables was recognized by the several plant protection organizations around the world. Currently, some 40 countries have approved one or more irradiated food items or groups of food

Irradiation as an alternative post harvest treatment

Energy Technology Data Exchange (ETDEWEB)

Satin, M [Agricultural Industries and Post-harvest Management Service, FAO, Rome (Italy); Loaharanu, P [Head, Food Preservation Section, Joint FAO/ IAEA Division of Nuclear Techniques in Food and Agriculture, Wagramerstr. 5, A-1400, Vienna (Austria)

1998-12-31

This current world population has significantly added to the pressures placed upon our finite resources and our resulting ability to feed ourselves. In order to cope with current and future demands, the two established lines of action, that is, reduced population growth and expansion of agricultural production, must be supplemented with the parallel activity of reducing food losses during and after harvest. For developing countries in particular, enormous post-harvest losses result from spillage, contamination, pests and physiological deterioration during storage. Studies in these countries indicate that post-harvest losses are enormous and amount to tens of millions of tons per year valued at billions of dollars. Programs to reduce post-harvest losses, if applied properly, can result in realistic yield increases between 10 and 30%, which can be directly converted into increased consumption for humans. Post-harvest losses vary greatly and are a function of the crop variety, pest combinations in the environment, climate, the system of harvesting, storage, handling, marketing, and even the social and cultural environment. Pests are among the most criticals of these factors. Because of the disastrous potential consequences of such pests, quarantine regulations prohibit the entrance of plants or products which might hide the unwanted pest from countries where it is known to exist. Quarantine treatments are can be chemical, physical or ionizing radiation treatment. Numerous investigations on the use of ionizing radiation for the disinfestation of fresh plant materials indicate that rather low dosages will control fruit-fly problems, thus making it well suited for quarantine treatment. The effectiveness of the irradiation as a broad spectrum quarantine treatment of fresh fruits and vegetables was recognized by the several plant protection organizations around the world. Currently, some 40 countries have approved one or more irradiated food items or groups of food

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

DEFF Research Database (Denmark)

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

2001-01-01

of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates theproblem 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...

Advanced Post-Irradiation Examination Capabilities Alternatives Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Jeff Bryan; Bill Landman; Porter Hill

2012-12-01

An alternatives analysis was performed for the Advanced Post-Irradiation Capabilities (APIEC) project in accordance with the U.S. Department of Energy (DOE) Order DOE O 413.3B, “Program and Project Management for the Acquisition of Capital Assets”. The Alternatives Analysis considered six major alternatives: ? No Action ? Modify Existing DOE Facilities – capabilities distributed among multiple locations ? Modify Existing DOE Facilities – capabilities consolidated at a few locations ? Construct New Facility ? Commercial Partnership ? International Partnerships Based on the alternatives analysis documented herein, it is recommended to DOE that the advanced post-irradiation examination capabilities be provided by a new facility constructed at the Materials and Fuels Complex at the Idaho National Laboratory.

Post-irradiation replication and repair in uv-irradiated cells of Proteus mirabilis depends on protein synthesis and a functioning rec/sup +/ gene

Energy Technology Data Exchange (ETDEWEB)

Hofemeister, J [Akademie der Wissenschaften der DDR, Gatersleben. Zentralinstitut fuer Genetik und Kulturpflanzenforschung

1977-02-28

The amount of and the molecular weight of newly synthesized DNA (piDNA) as well as its repair after uv irradiation in excision-proficient strains of P.mirabilis and E.coli K12 have been compared. A fraction of post-replication repair (PRR) in P.mirabilis is found to be dependent on de novo protein synthesis after uv irradiation. Pre-irradiation by uv and pre-treatment with nalidixic acid increase the efficiency of post-irradiation replication and PRR even in the presence of chloramphenicol. An inducible repair function in P.mirabilis is supposed to stimulate post-irradiation replication and repair.

Studies on Post-Irradiation DNA Degradation in Micrococcus Radiodurans, Strain RII51

DEFF Research Database (Denmark)

Auda, H.; Emborg, C.

1973-01-01

The influence of irradiation condition on post-irradiation DNA degradation was studied in a radiation resistant mutant of M. radiodurans, strain ${\\rm R}_{{\\rm II}}5$. After irradiation with 1 Mrad or higher more DNA is degraded in cells irradiated in wet condition than in cells irradiated with t...

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 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.

Influence of oxygen impurity atoms on defect clusters and radiation hardening in neutron-irradiated vanadium

International Nuclear Information System (INIS)

Bajaj, R.; Wechsler, M.S.

1975-01-01

Single crystal TEM samples and polycrystalline tensile samples of vanadium containing 60-640 wt ppm oxygen were irradiated at about 100 0 C to about 1.3 x 10 19 neutrons/cm 2 (E greater than 1 MeV) and post-irradiation annealed up to 800 0 C. The defect cluster density increased and the average size decreased with increasing oxygen concentration. Higher oxygen concentrations caused the radiation hardening and radiation-anneal hardening to increase. The observations are consistent with the nucleation of defect clusters by small oxygen or oxygen-point defect complexes and the trapping of oxygen at defect clusters upon post-irradiation annealing

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.)

Post harvest changes gamma-irradiated banana Prata

International Nuclear Information System (INIS)

Vilas Boas, E.V. de; Chitarra, A.B.; Chitarra, M.I.F.

1996-01-01

The effect of the gamma-irradiation was evaluated at 0.25 and 0.50 kGy, on the development of peel coloration, CO 2 and ethylene evolution, conversion of starch to sugars, pulp-to-peel ratio, pectic solubilization and activities of enzymes of the cell wall, pectin methylesterase (PME), and polygalacturonase (PG), during maturation of 'Prata' bananas. The gamma-irradiation did not affect the normal colour development of the fruits. An increase in the ethylene peak and a decrease in the CO 2 peak was observed. The gamma-irradiation did not affect the degradation of starch, while a delay in soluble sugar accumulation was noted on the 6 and 7 colour grades. The fruits subjected to 0.25 kGy had the highest increase in the pulp-to-peel relation, beginning with colour grade 5, due to a possible stress effect of that dose. An increase of pectin solubilization was observed. Higher PME activities were exhibited by irradiated fruits, although the gamma-irradiation suppressed the PG activity throughout the maturation period. The gamma-irradiation did not extend the post-harvest life of 'Prata' bananas. (author) [pt

Effect of Local Post Weld Heat Treatment on Tensile Properties in Friction Stir Welded 2219-O Al Alloy

Science.gov (United States)

Chu, Guannan; Sun, Lei; Lin, Caiyuan; Lin, Yanli

2017-11-01

To improve the formability of the aluminum alloy welds and overcome the size limitation of the bulk post weld heat treatment (BPWHT) on large size friction stir welded joints, a local post weld heat treatment method (LPWHT) was proposed. In this method, the resistance heating as the moving heat source is adopted to only heat the weld seam. The temperature field of LPWHT and its influence on the mechanical properties and formability of FSW 2219-O Al alloy joints was investigated. The evaluation of the tensile properties of FSW samples was also examined by mapping the global and local strain distribution using the digital image correlation methodology. The results indicated that the formability was improved greatly after LPWHT, while the hardness distribution of the FSW joint was homogenized. The maximum elongation can reach 1.4 times that of as-welded joints with increase the strength and the strain of the nugget zone increased from 3 to 8% when annealing at 300 °C. The heterogeneity on the tensile deformation of the as-welded joints was improved by the nugget zone showing large local strain value and the reason was given according to the dimple fracture characteristics at different annealing temperatures. The tensile strength and elongation of LPWHT can reach 93.3 and 96.1% of the BPWHT, respectively. Thus, the LPWHT can be advantageous compared to the BPWHT for large size welds.

Post-irradiation annealing of coarse-grained model alloys

Energy Technology Data Exchange (ETDEWEB)

Ray, P H.N.; Wilson, C; McElroy, R J [AEA Reactor Services, Harwell (United Kingdom)

1994-12-31

Thermal ageing and irradiation studies have been carried out on three model alloys (JPC, JPB, JPG) that have identical compositions except for different levels of phosphorus and/or copper. They have been irradiated in three conditions, as-received, heat treated to produce a coarse grained microstructure (similar to heat-affected-zone), and in this condition further aged at 450 C to produce a temper embrittled condition. One of the alloy have been subject to a post-irradiation anneal. The effect of these treatments on mechanical property changes has been characterized by Charpy testing and Vickers hardness measurements; the phosphorus segregation has been studied by a combination of STEM and Auger techniques.

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

Effect of post weld heat treatment on tensile properties and microstructure characteristics of friction stir welded armour grade AA7075-T651 aluminium alloy

Directory of Open Access Journals (Sweden)

P. Sivaraj

2014-03-01

Full Text Available This paper reports the effects of post weld heat treatments, namely artificial ageing and solution treatment followed by artificial ageing, on microstructure and mechanical properties of 12 mm thick friction stir welded joints of precipitation hardenable high strength armour grade AA7075-T651 aluminium alloy. The tensile properties, such as yield strength, tensile strength, elongation and notch tensile strength, are evaluated and correlated with the microhardness and microstructural features. The scanning electron microscope is used to characterie the fracture surfaces. The solution treatment followed by ageing heat treatment cycle is found to be marginally beneficial in improving the tensile properties of friction stir welds of AA7075-T651 aluminium alloy.

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

Effect of pulsed current and post weld aging treatment on tensile properties of argon arc welded high strength aluminium alloy

International Nuclear Information System (INIS)

Balasubramanian, V.; Ravisankar, V.; Reddy, G. Madhusudhan

2007-01-01

This paper reveals the effect of pulsed current and post weld aging treatment on tensile properties of argon arc welded AA7075 aluminium alloy. This alloy has gathered wide acceptance in the fabrication of light weight structures requiring high strength-to-weight ratio, such as transportable bridge girders, military vehicles, road tankers and railway transport systems. The preferred welding processes of high strength aluminium alloy are frequently gas tungsten arc welding (GTAW) process and gas metal arc welding (GMAW) process due to their comparatively easier applicability and better economy. Weld fusion zones typically exhibit coarse columnar grains because of the prevailing thermal conditions during weld metal solidification. This often results inferior weld mechanical properties and poor resistance to hot cracking. In this investigation, an attempt has been made to refine the fusion zone grains by applying pulsed current welding technique. Four different welding techniques have been used to fabricate the joints and they are: (i) continuous current GTAW (CCGTAW), (ii) pulsed current GTAW (PCGTAW), (iii) continuous current GMAW (CCGMAW) and (iv) pulsed current GMAW (PCGMAW) processes. As welded joint strength is much lower than the base metal strength and hence, a simple aging treatment has been given to improve the tensile strength of the joints. Current pulsing leads to relatively finer and more equi-axed grain structure in GTA and GMA welds. In contrast, conventional continuous current welding resulted in predominantly columnar grain structures. Post weld aging treatment is accompanied by an increase in tensile strength and tensile ductility

Effect of ultraviolet light irradiation on bond strength of fiber post: Evaluation of surface characteristic and bonded area of fiber post with resin cement

OpenAIRE

Reza, Fazal; Ibrahim, Nur Sukainah

2015-01-01

Objective: Fiber post is cemented to a root canal to restore coronal tooth structure. This research aims to evaluate the effect of ultraviolet (UV) irradiation on bond strength of fiber post with resin cement. Materials and Methods: A total of 40 of the two types of fiber posts, namely, FRC Prostec (FRC) and Fiber KOR (KOR), were used for the experiment. UV irradiation was applied on top of the fiber post surface for 0, 15, 20, and 30 min. The irradiated surface of the fiber posts (n = 5) wer...

Post-irradiation effects in CMOS integrated circuits

International Nuclear Information System (INIS)

Zietlow, T.C.; Barnes, C.E.; Morse, T.C.; Grusynski, J.S.; Nakamura, K.; Amram, A.; Wilson, K.T.

1988-01-01

The post-irradiation response of CMOS integrated circuits from three vendors has been measured as a function of temperature and irradiation bias. The author's have found that a worst-case anneal temperature for rebound testing is highly process dependent. At an anneal temperature of 80 0 C, the timing parameters of a 16K SRAM from vendor A quickly saturate at maximum values, and display no further changes at this temperature. At higher temperature, evidence for the anneal of interface state charge is observed. Dynamic bias during irradiation results in the same saturation value for the timing parameters, but the anneal time required to reach this value is longer. CMOS/SOS integrated circuits (vendor B) were also examined, and showed similar behavior, except that the saturation value for the timing parameters was stable up to 105 0 C. After irradiation to 10 Mrad(Si), a 16K SRAM (vendor C) was annealed at 80 0 C. In contrast to the results from the vendor A SRAM, the access time decreased toward prerad values during the anneal. Another part irradiated in the same manner but annealed at room temperature showed a slight increase during the anneal

Post retention and post/core shear bond strength of four post systems.

Science.gov (United States)

Stockton, L W; Williams, P T; Clarke, C T

2000-01-01

As clinicians we continue to search for a post system which will give us maximum retention while maximizing resistance to root fracture. The introduction of several new post systems, with claims of high retentive and resistance to root fracture values, require that independent studies be performed to evaluate these claims. This study tested the tensile and shear dislodgment forces of four post designs that were luted into roots 10 mm apical of the CEJ. The Para Post Plus (P1) is a parallel-sided, passive design; the Para Post XT (P2) is a combination active/passive design; the Flexi-Post (F1) and the Flexi-Flange (F2) are active post designs. All systems tested were stainless steel. This study compared the test results of the four post designs for tensile and shear dislodgment. All mounted samples were loaded in tension until failure occurred. The tensile load was applied parallel to the long axis of the root, while the shear load was applied at 450 to the long axis of the root. The Flexi-Post (F1) was significantly different from the other three in the tensile test, however, the Para Post XT (P2) was significantly different to the other three in the shear test and had a better probability for survival in the Kaplan-Meier survival function test. Based on the results of this study, our recommendation is for the Para Post XT (P2).

Proteomics of post-irradiation recovery in D. radiodurans

International Nuclear Information System (INIS)

Basu, Bhakti; Apte, Shree Kumar

2012-01-01

An extremophile Deinococcus radiodurans is bestowed with an extraordinary DNA repair ability that renders it virtually resistant to all known forms of DNA damage caused by ionizing radiations (10 kGy of gamma rays), UV (1 kJ/m 2 ) or weeks of desiccation etc. The genome of D. radiodurans encodes a unique combination of DNA repair pathways such as prokaryotic type RecFOR mediated homologous recombination (HR) and nucleotide/base excision repair along with eukaryotic type strand annealing (SA) and non-homologous end joining (NHEJ), but is devoid of universal prokaryotic DNA repair pathways such as RecBCD mediated HR, photo-reactivation and SOS response. Collective evidence obtained so far from multiple approaches, have indicated (i) that all genes essential for DNA repair are not necessarily induced following radiation stress (ii) early RecA independent DNA assembly occurs, and (iii) absolute necessity of RecA dependent HR for final genome restitution. The 6 kGy gamma irradiation inducible proteome dynamics were mapped during the post-irradiation growth arrest phase by 2D protein electrophoresis coupled with mass spectrometry. Radiation inducible expression of at least 33 proteins was evident in the first 1h of post irradiation recovery

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

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

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

Heat-treated mineral-yeast as a potent post-irradiation radioprotector

International Nuclear Information System (INIS)

Anzai, Kazunori; Ueno, Megumi; Nyui, Minako; Ikota, Nobuo; Kagiya, Tsutomu V.

2008-01-01

In vivo radioprotection of C3H mice by i.p. administration of Zn-, Mn-, Cu-, or Se-containing heat-treated Saccharomyces serevisiae yeast sample was examined. The 30-day survival of the group treated 30 min before 7.5 Gy whole-body X-irradiation with mineral-containing yeast powders suspended in 0.5% methylcellulose was significantly higher than that of control group. When mineral-yeast was administered immediately after irradiation, the survival rate was even higher and Zn- or Cu-yeast showed the highest rate (more than 90%). Although treatment with simple yeast showed a high survival rate (73%), it was significantly lower than that obtained by the Zn-yeast treatment. The effects of Zn-yeast were studied further. When the interval between irradiation and administration was varied, the protective activity of Zn-yeast decreased gradually by increasing the interval but was still significantly high for the administration at 10 h post-irradiation. The dose reduction factor of Zn-yeast (100 mg/kg, i.p. administration immediately after irradiation) was about 1.2. When the suspension of Zn-yeast was fractionated by centrifugation, the insoluble fraction showed a potent effect, while the soluble fraction had only a moderate effect. In conclusion, mineral-yeast, especially Zn-yeast, provides remarkable post-irradiation protection against lethal whole body X-irradiation. The activity is mainly attributable to the insoluble fraction, whereas some soluble components might contribute to the additional protective activity. (author)

Post irradiation effects (PIE) in integrated circuits

International Nuclear Information System (INIS)

Barnes, C.E.; Shaw, D.C.; Fleetwood, D.M.; Winokur, P.S.

1992-01-01

Post Irradiation Effects (PIE) ranging from normal recovery catastrophic failure have been observed in integrated circuits during the PIE period. These variations indicate that a rebound or PIE recipe used for radiation hardness assurance must be chosen with care. In this paper, the authors provide examples of PIE in a variety of integrated circuits of importance to spacecraft electronics

Joint research centre fusion materials irradiations in HFR: Present status and prospectives

International Nuclear Information System (INIS)

Casini, G.; Fenici, P.

1989-01-01

First a review is made of the Joint Research Centre experimental activity at HFR-Petten in the frame of the Fusion Technology and Safety Programme. The materials under investigation are: Cr-Ni Austenitic steels (316-L type) and Cr-Mn Austenitic steels (AMCR and FI type) as structural materials and Pb-17Li eutetic as tritium breeding material. The experiments on structural materials comprise: Sample irradiations with post-irradiation tensile tests (FRUST) Sample irradiations under constant load and post-irradiation strain measurement (TRIESTE) On-line creep tests (CRISP). The experiments on Pb-17Li breeder material regard sample irradiations to investigate tritium production and recovery as well as tritium permeation through blanket structures (LIBRETTO Experiment). Both irradiations on structural and breeding materials will be pursued up to the end of the current JRC-Multiannual Programme (1988-1991) and even further. In the last part of the paper expected developments of the testing programme at HFR are discussed. New areas of research should involve materials for divertor applications (NET/ITER) and advanced low activation composite materials for Commercial Power Reactors

Post-irradiation brain-necrosis resulting in apoplexia and death after 33 years of irradiation

Energy Technology Data Exchange (ETDEWEB)

Froehlich, A [Foevarosi Laszlo Korhaz, Budapest (Hungary). Korbonctani es Korszoevettani Oszt.

1980-04-01

A case of post-irradiation brain-necrosis resulting in apoplexia of the cerebellum after 33 years of irradiation (19984 r.) of a presumptive cerebellar tumour is reported. The pathohistologic study revealed symptoms of the ''late'' damage and the vascular changes appeared to be the most prominent. The thickening of the vessel walls, hyperplasia of collagen fibres and deposition of calcium in the media, were the most characteristic lesions revealed. In some of the small vessels isolated calcification of the media was observed. It seems most probable that in the development of apoplexia vascular alterations could play an important role. In the available literature no report has been found on a similarly long interval elapsing between the irradiation and death.

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

Strand breaks, base release and post-irradiation changes in DNA γ-irradiated in dilute O2-saturated aqueous solution

International Nuclear Information System (INIS)

Ward, J.F.; Kuo, I.

1976-01-01

Gamma irradiation of DNA in dilute O 2 -saturated aqueous solution releases free bases and damaged bases from the macromolecule. The yields of these products were measured after column chromatographic separation. For double stranded DNA the immediate yield of bases varies from G = 0.012 for cytosine to G = 0.033 for adenine. The yields of released bases increase with post-irradiation time (the majority of the increase occurs in the first 2 hrs.) to yields in the range of G = 0.07 +- 0.012. Yields of two released damaged thymine radiation products from γ-irradiated 3 H thymine labelled DNA also increased with post-irradiation time. Strand breaks were measured in γ-irradiated single stranded DNA the initial yield G = 0.02 was low but increased with time to G = 0.07. No direct correlation between strand-break production and release of low molecular weight products is possible. The findings are discussed in terms of damage to DNA in vivo and its enzymatic repair

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1982-01-01

The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

Shear Wave Elastography--A New Quantitative Assessment of Post-Irradiation Neck Fibrosis.

Science.gov (United States)

Liu, K H; Bhatia, K; Chu, W; He, L T; Leung, S F; Ahuja, A T

2015-08-01

Shear wave elastography (SWE) is a new technique which provides quantitative assessment of soft tissue stiffness. The aim of this study was to assess the reliability of SWE stiffness measurements and its usefulness in evaluating post-irradiation neck fibrosis. 50 subjects (25 patients with previous radiotherapy to the neck and 25 sex and age-matched controls) were recruited for comparison of SWE stiffness measurements (Aixplorer, Supersonic Imagine). 30 subjects (16 healthy individuals and 14 post-irradiated patients) were recruited for a reliability study of SWE stiffness measurements. SWE stiffness measurements of the sternocleidomastoid muscle and the overlying subcutaneous tissues of the neck were made. The cross-sectional area and thickness of the sternocleidomastoid muscle and the overlying subcutaneous tissue thickness of the neck were also measured. The post-irradiation duration of the patients was recorded. The intraclass correlation coefficients for the intraoperator and interoperator reliability of deep and subcutaneous tissue SWE stiffness ranged from 0.90-0.99 and 0.77-0.94, respectively. The SWE stiffness measurements (mean +/- SD) of deep and subcutaneous tissues were significantly higher in the post-irradiated patients (64.6 ± 46.8 kPa and 63.9 ± 53.1 kPa, respectively) than the sex and age-matched controls (19.9 ± 7.8 kPa and 15.3 ± 8.37 respectively) (p < 0.001). The SWE stiffness increased with increasing post-irradiation therapy duration in the Kruskal Wallis test (p < 0.001) and correlated with muscle atrophy and subcutaneous tissue thinning (p < 0.01). SWE is a reliable technique and may potentially be an objective and specific tool in quantifying deep and subcutaneous tissue stiffness, which in turn reflects the severity of neck fibrosis. © Georg Thieme Verlag KG Stuttgart · New York.

Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

International Nuclear Information System (INIS)

Bratton, R.L.

1997-05-01

As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

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

Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

Energy Technology Data Exchange (ETDEWEB)

Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

Post-irradiation examinations of THERMHET composite fuels for transmutation

Science.gov (United States)

Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

2003-07-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

Post-irradiation examinations of THERMHET composite fuels for transmutation

Energy Technology Data Exchange (ETDEWEB)

Noirot, J. E-mail: jnoirot@cea.fr; Desgranges, L.; Chauvin, N.; Georgenthum, V

2003-07-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl{sub 2}O{sub 4} spinel inert matrix and around 40% weight of UO{sub 2} to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

Post-irradiation examinations of THERMHET composite fuels for transmutation

International Nuclear Information System (INIS)

Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

2003-01-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2 O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour

Post-irradiation brain-necrosis resulting in apoplexia and death after 33 years of irradiation

International Nuclear Information System (INIS)

Froehlich, A.

1980-01-01

A case of post-irradiation brain-necrosis resulting in apoplexia of the cerebellum after 33 years of irradiation (19984 r.) of a presumptive cerebellar tumour is reported. The pathohistologic study revealed symptoms of the ''late'' damage and the vascular changes appeared to be the most prominent. The thickening of the vessel walls, hyperplasia of collagen fibres and deposition of calcium in the media, were the most characteristic lesions revealed. In some of the small vessels isolated calcification of the media was observed. It seems most probable that in the development of apoplexia vascular alterations could play an important role. In the available literature no report has been found on a similarly long interval elapsing between the irradiation and death. (author)

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.

Post harvest controlling of orchid thrips on cut flowers by irradiation

International Nuclear Information System (INIS)

Bansiddhi, K.; Siriphontangmun, S.

1999-01-01

Post-harvest controlling of orchid thrips, Thrips palmi Karny on cut flowers by irradiation was conducted during October 1992 to September 1997 at the Thai Irradiation Centre (TIC) and Division of Entomology and Zoology, Department of Agriculture, Thailand. The studies were carried out by conducting experiments on irradiation of cut flowers for controlling thrips with doses ranging from 0.1 to 1 kGy. The vase-life of radiated cut flowers was evaluated. Colonies of thrips were established in the laboratory in order to determine radiation sensitivity of various development stages of thrips and also to assess the occurrence of natural infestations by examining commercial market quality flowers from growers where management practices can be identified. Results from five years of research on post harvest control of thrips on orchids and cut flowers by irradiation showed that despite intensive investigation, difficulty in permanent establishment of a laboratory colony of Thrips palmi Karny for bioassays continued. The snap bean rearing method for rearing large number of thrips has bean developed, although it is less satisfactory than desirable. It has given sufficient numbers for testing in the 6th experiment. The maximum dose tolerated by Dendrobium orchid flowers at ambient temperature (25-30 deg. C) was below 0.5 kGy, but at a pre- and post irradiation temperature 15-18 deg. C, the maximum dose tolerated approached 0.75-0.8 kGy. The effective dose for control Thrips palmi Karny, however, was higher than 0.75 kGy. (author)

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.

The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

International Nuclear Information System (INIS)

White, R.J.

2001-01-01

Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

International Nuclear Information System (INIS)

Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.P.; Viehrig, H.W.

1992-01-01

This work deals with the mechanical properties of RPV steels used WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy V-notch specimens and tensile test specimens were irradiated in the WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 x 10 18 and 5 x 10 19 n/cm 2 (E>1MeV). Post-irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. (orig.)

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

International Nuclear Information System (INIS)

Was, Gary

2017-01-01

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

Energy Technology Data Exchange (ETDEWEB)

Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

2017-06-02

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

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

Post irradiation examinations cooperation and worldwide utilization of facilities

International Nuclear Information System (INIS)

Karlsson, Mikael

2009-01-01

Status of post irradiation examinations in Studsvik's facilities, cooperation and worldwide utilization of facilities, was described. Studsvik cooperate with irradiation facilities, as Halden, CEA and JAEA, as well as other hot cell facilities (examples, PSI, ITU and NFD) universities (example, the Royal Institute of Technology in Sweden) in order to be able to provide everything asked for by the nuclear community. Worldwide cooperation for effective use of expensive and highly specialized facilities is important, and the necessity of cooperation will be more and more recognized in the future. (author)

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

Effect of gamma irradiation on the microstructure and post-mortem anaerobic metabolism of bovine muscle

International Nuclear Information System (INIS)

Yook, H.-S.; Lee, J.-W.; Lee, K.-H.; Kim, M.-K.; Song, C.-W.; Byun, M.-W.

2001-01-01

Experiments were performed to study the effect of gamma irradiation on morphological properties and post-mortem metabolism in bovine M. sternomandibularis with special reference to ultrastructure, shear force, pH and ATP breakdown. The shortening of sarcomere was not observed in gamma-irradiated muscle, however, the disappearance of M-line and of A- and I-bands was perceptible. During cold storage, the destruction of muscle bundles was faster in the gamma-irradiated muscle than in the non-irradiated with a dose-dependent manner. The same is true for the post mortem pH drop and ATP breakdown. So, experimental results confirmed that the anaerobic metabolism and morphological properties are noticeably affected by gamma irradiation in beef

Clinical and pathologic factors predictive of biochemical control following post-prostatectomy irradiation

International Nuclear Information System (INIS)

Stromberg, Jannifer S.; Ziaja, Ellen L.; Horwitz, Eric M.; Vicini, Frank A.; Brabbins, Donald S.; Dmuchowski, Carl F.; Gonzalez, Jose; Martinez, Alvaro A.

1996-01-01

Purpose/Objective: Indications for post-prostatectomy radiation therapy are not well defined. We reviewed our experience treating post-prostatectomy patients with external beam irradiation to assess clinical and pathologic factors predictive of biochemical control. Materials and Methods: Between 1/87 and 3/93, 61 patients received post-operative tumor bed irradiation with a median dose of 59.4 Gy (50.4 - 68 Gy). Median follow-up was 4.1 years (7.6 months - 8.3 years) from irradiation. Patients were treated for the following reasons: 1) adjuvantly, within 6 months of surgery for extracapsular extension, seminal vesicle involvement, or positive surgical margins (n=38); 2) persistently elevated PSA post-operatively (n=2); 3) rising PSA >6 months after surgery (n=9); and 4) biopsy proven local recurrence (n=12). No patients had known nodal or metastatic disease. All patients had post-radiation PSA data available. Biochemical control was the endpoint studied using Kaplan-Meier life table analysis. Biochemical control was defined as the ability to maintain an undetectable PSA ( 4 and ≤1 0, >10 and ≤20, and > 20 ng/ml. The 3 year actuarial rates of biochemical control were 100% for group 1, 66.7% for group 2, 61.5% for group 3, and 28.6% for group 4. Pre-RT PSA values were also evaluated. Univariate Cox models indicated lower presurgical and pre-RT PSA values were predictive of biochemical control (p=0.017, p 6 months after surgery (group 3), the 3 year actuarial rate of biochemical control was 55.6%. The 3 year actuarial rate of biochemical control for patients treated for a biopsy proven recurrence (group 4) was 8.3%. By pair-wise log rank test, the rates of biochemical control were significantly different between groups 1 and 3 (p=0.036), groups 1 and 4 (p<0.001), and groups 3 and 4 (p=0.009). Conclusion: Biochemical control was achieved in approximately half of the patients treated with post-operative prostatic fossa irradiation. Elevated presurgical and pre

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

Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

2007-03-01

A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Microstructural basis and crack growth theories for post-irradiation ductility loss in Nimonic PE16

International Nuclear Information System (INIS)

Chang, A.L.

1982-01-01

A study has been carried out to investigate the degradation of postirradiation ductility at reactor temperatures in Nimonic PE16, a Fe-Cr-Ni-based precipitation-hardened superalloy. Fractographic and microstructural investigations show that the grain matrix is capable of deformation and does not limit the postirradiation tensile ductility. Grain-boundary helium bubbles formed during neutron irradiation seem to be crack nucleation sites under stress. Growth and coalescence of these microcracks under stress lead to intergranular fracture. A rigid-grain fracture model is shown to be able to correlate the observed microstructures with most features of the mechanical properties, except the strain rate dependence of the ductility. By incorporating the interactions between diffusion and plastic deformation, a plastic-grain fracture model has been developed which can explain all postirradiation tensile ductility data quantitatively. 13 references

Post Irradiation Capabilities at the Idaho National Laboratory

International Nuclear Information System (INIS)

Schulthess, J.L.; Rosenberg, K.E.

2011-01-01

The U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) oversees the efforts to ensure nuclear energy remains a viable option for the United States. A significant portion of these efforts are related to post-irradiation examinations (PIE) of highly activated fuel and materials that are subject to the extreme environment inside a nuclear reactor. As the lead national laboratory, Idaho National Laboratory (INL) has a rich history, experience, workforce and capabilities for performing PIE. However, new advances in tools and techniques for performing PIE now enable understanding the performance of fuels and materials at the nano-scale and smaller level. Examination at this level is critical since this is the scale at which irradiation damage occurs. The INL is on course to adopt these advanced tools and techniques to develop a comprehensive nuclear fuels and materials characterization capability that is unique in the world. Because INL has extensive PIE capabilities currently in place, a strong foundation exist to build upon as new capabilities are implemented and work load increases. In the recent past, INL has adopted significant capability to perform advanced PIE characterization. Looking forward, INL is planning for the addition of two facilities that will be built to meet the stringent demands of advanced tools and techniques for highly activated fuels and materials characterization. Dubbed the Irradiated Materials Characterization Laboratory (IMCL) and Advanced Post Irradiation Examination Capability, these facilities are next generation PIE laboratories designed to perform the work of PIE that cannot be performed in current DOE facilities. In addition to physical capabilities, INL has recently added two significant contributors to the Advanced Test Reactor-National Scientific User Facility (ATR-NSUF), Oak Ridge National Laboratory and University of California, Berkeley.

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

Effect of ultraviolet light irradiation period on bond strengths between fiber-reinforced composite post and core build-up composite resin.

Science.gov (United States)

Asakawa, Yuya; Takahashi, Hidekazu; Iwasaki, Naohiko; Kobayashi, Masahiro

2014-01-01

The aim of the present study was to characterize the effects of the ultraviolet light (UV) irradiation period on the bond strength of fiber-reinforced composite (FRC) posts to core build-up resin. Three types of FRC posts were prepared using polymethyl methacrylate, urethane dimethacrylate, and epoxy resin. The surfaces of these posts were treated using UV irradiation at a distance of 15 mm for 0 to 600 s. The pull-out bond strength was measured and analyzed with the Dunnett's comparison test (α=0.05). The bond strengths of the post surfaces without irradiation were 6.9 to 7.4 MPa; those after irradiation were 4.2 to 26.1 MPa. The bond strengths significantly increased after 15 to 120-s irradiation. UV irradiation on the FRC posts improved the bond strengths between the FRC posts and core build-up resin regardless of the type of matrix resin.

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

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

Energy Technology Data Exchange (ETDEWEB)

Magielsen, A.J., E-mail: magielsen@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Jong, M.; Bakker, T.; Luzginova, N.V.; Mutnuru, R.K.; Ketema, D.J.; Fedorov, A.V. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands)

2011-08-31

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 deg. C and 500 deg. C. During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 deg. C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 deg. C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.

Post-irradiation hyperamylasemia as a biological dosimeter

International Nuclear Information System (INIS)

Dubray, Bernard; Girinski, Theo; Hennequin, Christophe; Socie, Gerard; Bonnay, Marc; Cosset, J.M.; Thames, H.D.; Becciolini, Aldo; Porciani, Sauro

1992-01-01

Serum alpha-amylasemia was measured before and 24 h after either total body (31 patients) or localized irradiation including salivary glands (40 patients) or pancreatic area (22 patients). A significant increase in amylasemia was observed for doses to the parotid glands larger than 0.5Gy. A sigmoid function of dose was fitted to the data and predicted a maximum amylasemia level for doses larger than 4Gy and smaller than 10Gy. Raw data from other published series were adequately described by the same model. However, confidence limits of parameters remained wide, because of a considerable interindividual variability. Post-irradiation hyperamylasemia appears to provide good criteria for triage of accidentally irradiated patients: 24 h after a dose larger than 2Gy to the parotid glands, 91% of patients had an amylasemia level higher than 2.5-fold the upper normal value (sensitivity). Conversely, 96% had their serum amylasemia lower than 2.5-fold the upper normal value when dose was smaller than 2Gy (specificity). However, a retrospective estimation of the absorbed dose (dosimetry) is not likely to be very precise because of the large interindividual variability. (author). 21 ref.; 1 fig.; 3 tabs

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

The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

International Nuclear Information System (INIS)

Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

1979-09-01

A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

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)

Detection of irradiation induced modifications in foodstuff DNA using 32p post-labelling

International Nuclear Information System (INIS)

Hoey, B.M.; Swallow, A.J.; Margison, G.P.

1991-01-01

DNA post-labelling has been used successfully to detect damage to DNA caused by a range of damaging agents. The assay results in a fingerprint of changes induced in DNA which might, in principle, be useful as a test for the detection of the irradiation of foods. The authors present their DNA extraction and 32 p post-labelling methods from chicken or cooked prawn samples and their analysis method (High Performance liquid chromatography). It's hoped that these results could form the basis of a test to detect if foods have been irradiated

Interaction of post harvest disease control treatments and gamma irradiation on mangoes

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.I.; Cooke, A.W. (Department of Primary Industries, Indooroopilly (Australia)); Boag, T.S. (Riverina-Murray Inst. of Higher Education, Wagga Wagga (Australia). School of Agriculture); Izard, M. (Australian Nuclear Science and Technology Organisation, Lucas Heights (Australia)); Panitz, M. (Committee of Direction of Fruit Marketing, Brisbane Markets (Australia)); Sangchote, S. (Kasetsart Univ., Bangkok (Thailand))

1990-04-01

The effects of gamma irradiation and disease control treatments on disease severity and post harvest quality of several mango cultivars were investigated. In mangoes cv. Kensington Pride, irradiation doses ranging from 300-1200 Gy reduced disease, but the level of control was not commercially acceptable. Hot benomyl immediately followed by irradiation provided effective control of anthracnose (Colletotrichum gloeosporioides) and stem end rot (Dothiorella dominicana) during short-term storage (15 days at 20degC). The effects of the two treatments were additive. Satisfactory disease control was achieved during long-term controlled atmosphere storage when mangoes were treated with hot benomyl followed by prochloraz and then irradiated. Effects of fungicide treatment and irradiation were additive. Fungicide, or irradiation treatments alone, were unsatisfactory. Irradiation of cv. Kensington Pride at doses in excess of 600 Gy caused unacceptable surface damage. (author).

Interaction of post harvest disease control treatments and gamma irradiation on mangoes

International Nuclear Information System (INIS)

Johnson, G.I.; Cooke, A.W.; Boag, T.S.; Panitz, M.; Sangchote, S.

1990-01-01

The effects of gamma irradiation and disease control treatments on disease severity and post harvest quality of several mango cultivars were investigated. In mangoes cv. Kensington Pride, irradiation doses ranging from 300-1200 Gy reduced disease, but the level of control was not commercially acceptable. Hot benomyl immediately followed by irradiation provided effective control of anthracnose (Colletotrichum gloeosporioides) and stem end rot (Dothiorella dominicana) during short-term storage (15 days at 20degC). The effects of the two treatments were additive. Satisfactory disease control was achieved during long-term controlled atmosphere storage when mangoes were treated with hot benomyl followed by prochloraz and then irradiated. Effects of fungicide treatment and irradiation were additive. Fungicide, or irradiation treatments alone, were unsatisfactory. Irradiation of cv. Kensington Pride at doses in excess of 600 Gy caused unacceptable surface damage. (author)

Effects of ablation energy and post-irradiation on the structure and properties of titanium dioxide nanomaterials

International Nuclear Information System (INIS)

Guillén, G. García; Shaji, S.; Palma, M. I. Mendivil; Avellaneda, D.; Castillo, G.A.; Roy, T.K. Das

2017-01-01

Highlights: • Highlights • TiO_2 nanomaterials were prepared by PLALM. • Characterized these nanomaterials using TEM, XPS, XRD, optical and luminescence measurements. • Morphology of these nanomaterials were dependent on ablation wavelength, fluence and post-irradiation time. • Laser post irradiation modified the size, morphology and structure of these TiO_2 nanomaterials. - Abstract: Nanomaterials of titanium oxide were prepared by pulsed laser ablation of a titanium metal target in distilled water. The ablation was performed at different laser energy (fluence) using a nanosecond pulsed Nd:YAG laser output of 1064 and 532 nm. A post-irradiation of titanium oxide nanocolloids obtained by ablation using 532 nm was carried out to explore its effects on the structure and properties. Analysis of morphology, crystalline phase, elemental composition, chemical state, optical and luminescent properties were performed using Transmission Electron Microscopy (TEM), X-Ray Diffraction (XRD), X-Ray Photoelectron Spectroscopy (XPS), UV–-vis absorption spectroscopy and room temperature photoluminescence spectroscopy. It was found that titanium oxide nanomaterial morphologies and optical properties were determined by ablation wavelength and fluence. Further, nanocolloids prepared by 532 nm ablation showed a crystalline phase change by laser post-irradiation. The results showed that pulsed laser ablation in liquid as well as post-irradiation were effective in modifying the final structure and properties of titanium oxide nanocolloids.

Effects of ablation energy and post-irradiation on the structure and properties of titanium dioxide nanomaterials

Energy Technology Data Exchange (ETDEWEB)

Guillén, G. García [Universidad Autónoma de Nuevo León, Facultad de Ingeniería Mecánica y Eléctrica, San Nicolás de los Garza, Nuevo León 66455, México (Mexico); Shaji, S., E-mail: sshajis@yahoo.com [Universidad Autónoma de Nuevo León, Facultad de Ingeniería Mecánica y Eléctrica, San Nicolás de los Garza, Nuevo León 66455, México (Mexico); Universidad Autónoma de Nuevo León-CIIDIT, Apodaca, Nuevo León, México (Mexico); Palma, M. I. Mendivil [Centro de Investigación en Materiales Avanzados (CIMAV), Unidad Monterrey, PIIT, Apodaca, Nuevo León, México (Mexico); Avellaneda, D.; Castillo, G.A.; Roy, T.K. Das [Universidad Autónoma de Nuevo León, Facultad de Ingeniería Mecánica y Eléctrica, San Nicolás de los Garza, Nuevo León 66455, México (Mexico); and others

2017-05-31

Highlights: • Highlights • TiO{sub 2} nanomaterials were prepared by PLALM. • Characterized these nanomaterials using TEM, XPS, XRD, optical and luminescence measurements. • Morphology of these nanomaterials were dependent on ablation wavelength, fluence and post-irradiation time. • Laser post irradiation modified the size, morphology and structure of these TiO{sub 2} nanomaterials. - Abstract: Nanomaterials of titanium oxide were prepared by pulsed laser ablation of a titanium metal target in distilled water. The ablation was performed at different laser energy (fluence) using a nanosecond pulsed Nd:YAG laser output of 1064 and 532 nm. A post-irradiation of titanium oxide nanocolloids obtained by ablation using 532 nm was carried out to explore its effects on the structure and properties. Analysis of morphology, crystalline phase, elemental composition, chemical state, optical and luminescent properties were performed using Transmission Electron Microscopy (TEM), X-Ray Diffraction (XRD), X-Ray Photoelectron Spectroscopy (XPS), UV–-vis absorption spectroscopy and room temperature photoluminescence spectroscopy. It was found that titanium oxide nanomaterial morphologies and optical properties were determined by ablation wavelength and fluence. Further, nanocolloids prepared by 532 nm ablation showed a crystalline phase change by laser post-irradiation. The results showed that pulsed laser ablation in liquid as well as post-irradiation were effective in modifying the final structure and properties of titanium oxide nanocolloids.

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

Defects annihilation behavior of neutron-irradiated SiC ceramics densified by liquid-phase-assisted method after post-irradiation annealing

Directory of Open Access Journals (Sweden)

Mohd Idzat Idris

2016-12-01

Full Text Available Numerous studies on the recovery behavior of neutron-irradiated high-purity SiC have shown that most of the defects present in it are annihilated by post-irradiation annealing, if the neutron fluence is less than 1×1026 n/m2 (>0.1MeV and the irradiation is performed at temperatures lower than 973K. However, the recovery behavior of SiC fabricated by the nanoinfiltrated and transient eutectic phase (NITE process is not well understood. In this study, the effects of secondary phases on the irradiation-related swelling and recovery behavior of monolithic NITE-SiC after post-irradiation annealing were studied. The NITE-SiC specimens were irradiated in the BR2 reactor at fluences of up to 2.0–2.5×1024 n/m2 (E>0.1MeV at 333–363K. This resulted in the specimens swelling up ∼1.3%, which is 0.1% higher than the increase seen in concurrently irradiated high-purity SiC. The recovery behaviors of the specimens after post-irradiation thermal annealing were examined using a precision dilatometer; the specimens were heated at temperatures of up to 1673K using a step-heating method. The recovery curves were analyzed using a first-order model, and the rate constants for each annealing step were obtained to determine the activation energy for volume recovery. The NITE-A specimen (containing 12 wt% sintering additives recovered completely after annealing at ∼1573K; however, it shrank because of the volatilization of the oxide phases at 1673K. The NITE-B specimen (containing 18wt% sintering additives did not recover fully, since the secondary phase (YAG was crystallized during the annealing process. The recovery mechanism of NITE-A SiC was based on the recombination of the C and Si Frenkel pairs, which were very closely sited or only slightly separated at temperatures lower than 1223K, as well as the recombination of the slightly separated C Frenkel pairs and the migration of C and Si interstitials at temperatures of 1223–1573K. That is to say, the

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

Modeling plastic deformation of post-irradiated copper micro-pillars

Energy Technology Data Exchange (ETDEWEB)

Crosby, Tamer, E-mail: tcrosby@ucla.edu; Po, Giacomo, E-mail: gpo@ucla.edu; Ghoniem, Nasr M., E-mail: ghoniem@ucla.edu

2014-12-15

We present here an application of a fundamentally new theoretical framework for description of the simultaneous evolution of radiation damage and plasticity that can describe both in situ and ex situ deformation of structural materials [1]. The theory is based on the variational principle of maximum entropy production rate; with constraints on dislocation climb motion that are imposed by point defect fluxes as a result of irradiation. The developed theory is implemented in a new computational code that facilitates the simulation of irradiated and unirradiated materials alike in a consistent fashion [2]. Discrete Dislocation Dynamics (DDD) computer simulations are presented here for irradiated fcc metals that address the phenomenon of dislocation channel formation in post-irradiated copper. The focus of the simulations is on the role of micro-pillar boundaries and the statistics of dislocation pinning by stacking-fault tetrahedra (SFTs) on the onset of dislocation channel and incipient surface crack formation. The simulations show that the spatial heterogeneity in the distribution of SFTs naturally leads to localized plastic deformation and incipient surface fracture of micro-pillars.

First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

International Nuclear Information System (INIS)

Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

2006-01-01

In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J. [Pacific Northwest National Laboratory, Richland, WA (United States)

1996-10-01

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

Post-irradiation inactivation, protection, and repair of the sulfhydryl enzyme malate synthase

International Nuclear Information System (INIS)

Durchschlag, H.; Zipper, P.

1985-01-01

Malate synthase from baker's yeast, a trimeric sulfhydryl enzyme with one essential sulfhydryl group per subunit, was inactivated by 2 kGy X-irradiation in air-saturated aqueous solution (enzyme concentration: 0.5 mg/ml). The radiation induced changes of enzymic activity were registered at about 0,30,60 h after irradiation. To elucidate the role of OH - , O 2 , and H 2 O 2 in the X-ray inactivation of the enzyme, experiments were performed in the absence of presence of different concentrations of specific additives (formate, superoxide dismutase, catalase). These additives were added to malate synthase solutions before or after X-irradiation. Moreover, repairs of inactivated malate synthase were initiated at about 0 or 30 h after irradiation by means of the sulfhydryl agent dithiothreitol. Experiments yielded the following results: 1. Irradiation of malate synthase in the absence of additives inactivated the enzyme immediately to a residual activity Asub(r)=3% (corresponding to a D 37 =0.6 kGy), and led to further slow inactivation in the post-irradiation phase. Repairs, initiated at different times after irradiation, restored enzymic activity considerably. The repair initiated at t=0 led to Asub(r)=21%; repairs started later on resulted in somewhat lower activities. The decay of reparability, however, was found to progress more slowly than post-irradiation inactivation itself. After completion of repair the activities of repaired samples did not decrease significantly. 2. The presence of specific additives during irradiation caused significant protective effects against primary inactivation. The protection by formate was very pronounced (e.g., Asub(r)=72% and D 37 =6 kGy for 100 mM formate). The presence of catalytic amounts of superoxide dismutase and/or catalase exhibited only minor effects, depending on the presence and concentration of formate. (orig.)

Hot cell facilities for post irradiation examination

International Nuclear Information System (INIS)

Mishra, Prerna; Bhandekar, Anil; Pandit, K.M.; Dhotre, M.P.; Rath, B.N.; Nagaraju, P.; Dubey, J.S.; Mallik, G.K.; Singh, J.L.

2017-01-01

Reliable performance of nuclear fuels and critical core components has a large bearing on the economics of nuclear power and radiation safety of plant operating personnel. In view of this, Post Irradiation Examination (PIE) is periodically carried out on fuels and components to generate feedback information which is used by the designers, fabricators and the reactor operators to bring about changes for improved performance of the fuel and components. Examination of the fuel bundles has to be carried out inside hot cells due to their high radioactivity

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

Science.gov (United States)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

A comparative study of post-irradiation growth kinetics of spheroids and monolayers

International Nuclear Information System (INIS)

Dertinger, J.; Luecke-Huhle, C.

1975-01-01

Post-irradiation growth kinetics of γ-irradiated spheroid and monolayer cells in exponential growth phase was investigated by means of dose-response curves based on cell counts after specified time intervals following irradiation. A mathematical model of cell-growth after irradiation was fitted to these curves. The model parameters (related to division delay and growth of non-surviving cells) obtained from this analysis consistently indicated increasing resistance to sub-lethal damage of cells cultured as multicellular spheroids under conditions of increasing three-dimensional contact. In contrast, no indication of an increased radiation-resistance was found with cells cultured on a substratum under a variety of conditions. (author)

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)

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

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)

Report of Post Irradiation Examination for Dry Process Fuel

International Nuclear Information System (INIS)

Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

2006-08-01

The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

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

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1995-01-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding-was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2 MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1997-08-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was.not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

Clinical analysis of post-irradiation sensorineural hearing loss in patients suffering from nasopharyngeal carcinoma

International Nuclear Information System (INIS)

Lu Xueguan; Liu Zhiyong; Zhang Liyuan; Tian Ye

2005-01-01

Objective: To investigate the incidence of post-irradiation sensorineural hearing loss (SNHL) in patients suffering from nasopharyngeal carcinoma and to evaluate its potentially contributing factors. Methods: Pure tonetest and impedance audiography were carried out in patients suffering from nasopharyngeal carcinoma with a post-irradiation follow-up time over 1 year. Additionally, the test results were combined with clinical data and analyzed retrospectively. Results: The follow-up time of all patients ranged from 12 to 94 months (median 53 months). The incidences of SNHL at low and high frequencies were 8% and 42% respectively. Univariate analysis showed that patient's age and follow-up time affected the incidence of SNHL at high frequencies (t=2.051, P=0.0269; t=2.978, P=0.0011), but sex, preirradiation subjective hearing loss, irradiation dose and chemotherapy including cisplatin had no significance. Multivariate analysis by Binary Logistic Regression revealed that the risk of SNHL was correlated with patient's age and follow-up time (P=0.02; P=0.009). Conclusion: Post-irradiation SNHL at high frequencies in patients suffering from nasopharyngeal carcinoma is more common than that at low frequencies. The independent prognostic factors for development of SNHL at high frequencies are patient's age and follow-up time. But the role of preirradiation hearing level ,irradiation dose and chemotherapy including cisplatin are not conclusive and further research is needed. (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

Design of a decontamination section of the post-irradiation examination laboratory

International Nuclear Information System (INIS)

Homberger, Victor; Coronel, Ruben R.; Laumann, Victor; Perez, Jorge O.; Quinteros, Andrea N.; Ratner, Marcos

1999-01-01

The Post-Irradiation Examination Laboratory activities include the decontamination of expensive equipment of different sizes and weight, involving the complexity and extension of the necessary decontamination. A decontamination section has been designed for that purpose. (author)

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)

Familial study of ataxia telangiectasia. Heterozygotes identification on the basis of sensitivity of gamma-irradiated cultures to caffeine post-treatment

International Nuclear Information System (INIS)

Pawlak, A.L.; Kotecki, M.

1994-01-01

The effects of caffeine (CF), γ-irradiation + CF post-treatment on chromosomal aberrations were studied in lymphocyte cultures from a patient with ataxia telangiectasia (AT), his parents and brother. In the studied family both the homozygotes and the obligatory heterozygotes of AT showed increased sensitivity to CF post-treatment. Individual differences in sensitivity to γ-irradiation + CF post-treatment proved to be correlated with the sensitivity of non-irradiated cells to CF treatment, but not to γ-irradiation. (author). 19 refs, 1 fig., 1 tab

Familial study of ataxia telangiectasia. Heterozygotes identification on the basis of sensitivity of gamma-irradiated cultures to caffeine post-treatment

Energy Technology Data Exchange (ETDEWEB)

Pawlak, A.L.; Kotecki, M. [Polska Akademia Nauk, Poznan (Poland). Zaklad Genetyki Czlowieka; Ignatowicz, R. [Centrum Zdrowia Dziecka, Warsaw (Poland)

1994-12-31

The effects of caffeine (CF), {gamma}-irradiation + CF post-treatment on chromosomal aberrations were studied in lymphocyte cultures from a patient with ataxia telangiectasia (AT), his parents and brother. In the studied family both the homozygotes and the obligatory heterozygotes of AT showed increased sensitivity to CF post-treatment. Individual differences in sensitivity to {gamma}-irradiation + CF post-treatment proved to be correlated with the sensitivity of non-irradiated cells to CF treatment, but not to {gamma}-irradiation. (author). 19 refs, 1 fig., 1 tab.

Enhancement of tumor cell killing in vitro by pre- and post-irradiation exposure to aclacinomycin A

International Nuclear Information System (INIS)

Bill, C.A.; Mendoza, A.; Vrdoljak, E.; Tofilon, P.J.

1993-01-01

Aclacinomycin A (ACM), a potent inducer of leukemic cell differentiation, significantly enhances the radiosensitivity of a human colon tumor cell line (Clone A) when cultures are exposed to 15-nM concentrations for 3 days before irradiation. We now demonstrate that incubation with ACM after irradiation can also enhance Clone A cell killing. The maximum increase in cell killing, based on colony-forming ability, occurred when Clone A cells were exposed for 1 h to 5 μM ACM model added 1 or 2 h after irradiation. The post-irradiation ACM protocol reduced the terminal slope (as reflected by D o ) of the radiation cell survival curve with no change in the low-dose, shoulder region of the curve (D q value). In contrast, for pre-irradiation treatment with ACM (15 nM, 3 days), the shoulder region of the curve was reduced with no change in the terminal slope. For pre- and post-irradiation ACM treatment the dose enhancement factors at 0.10 survival were 1.22 and 1.28, respectively. When ACM was given both before and after irradiation both the shoulder and terminal slope values decreased to produce a dose enhancement factor at a surviving fraction of 0.10 of 1.50. These data suggest that the enhanced cell killing produced by pre- and post-irradiation treatment with ACM is achieved through different mechanisms. (author) 26 refs., 3 tabs., 2 figs

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1981-04-01

Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

Mathematical Model for Post-Irradiation Haemopoiesis

Energy Technology Data Exchange (ETDEWEB)

Okunewick, J. P.; Kretchmar, A. L. [Rand Corporation, Santa Monica, CA (United States); Medical Division, Oak Ridge Associated Universities, Oak Ridge, TN (United States)

1968-08-15

A model for haemopoiesis has been constructed based on the following hypothesis: (a) Haemopoietic stem cells have the capability of either reproducing as stem cells or differentiating into specialized blood cells of at least two different types; (b) The size of the stem-cell compartment is in part regulated by the rate of increase due to stem-cell reproduction and in part by the rate of loss of stem cells through differentiation; (c) In addition, the size of the stem-cell compartment is in part regulated by a competitive cell-to-cell interaction between the stem-cells themselves and between the differentiating cells and the stem-cells, such that the presence of an exceptionally large number of either cell type would have a repressive effect on the rate of increase of the stem-cell population. This model has been applied to the post-irradiation erythropoietic behaviour of the rat. In the computer studies with the model, an X-ray dose sufficient to inhibit reproduction in 50% of the erythroid stem cells was assumed. It was also assumed that reproduction and differentiation are genetically separately controlled processes and that, therefore, some part of the reproductively injured cells were still capable of differentiation. Under these conditions the model predicted an abortive rise in reticulocyte number, peaking at about 6 days. True recovery was predicted to occur at about 16 days. Both the abortive rise and the true recovery were also present in those segments of the model representing earlier erythroid cells, occurring at progressively earlier times in progressively more primitive cells. Comparison of the model's predictions with experimentally obtained data for post-irradiation erythroid recovery showed a good agreement both with respect to the time of the abortive peak and the time of true recovery. (author)

Mathematical Model for Post-Irradiation Haemopoiesis

International Nuclear Information System (INIS)

Okunewick, J.P.; Kretchmar, A.L.

1968-01-01

A model for haemopoiesis has been constructed based on the following hypothesis: (a) Haemopoietic stem cells have the capability of either reproducing as stem cells or differentiating into specialized blood cells of at least two different types; (b) The size of the stem-cell compartment is in part regulated by the rate of increase due to stem-cell reproduction and in part by the rate of loss of stem cells through differentiation; (c) In addition, the size of the stem-cell compartment is in part regulated by a competitive cell-to-cell interaction between the stem-cells themselves and between the differentiating cells and the stem-cells, such that the presence of an exceptionally large number of either cell type would have a repressive effect on the rate of increase of the stem-cell population. This model has been applied to the post-irradiation erythropoietic behaviour of the rat. In the computer studies with the model, an X-ray dose sufficient to inhibit reproduction in 50% of the erythroid stem cells was assumed. It was also assumed that reproduction and differentiation are genetically separately controlled processes and that, therefore, some part of the reproductively injured cells were still capable of differentiation. Under these conditions the model predicted an abortive rise in reticulocyte number, peaking at about 6 days. True recovery was predicted to occur at about 16 days. Both the abortive rise and the true recovery were also present in those segments of the model representing earlier erythroid cells, occurring at progressively earlier times in progressively more primitive cells. Comparison of the model's predictions with experimentally obtained data for post-irradiation erythroid recovery showed a good agreement both with respect to the time of the abortive peak and the time of true recovery. (author)

Effect of irradiation on the post-harvest life of potatoes

International Nuclear Information System (INIS)

Mahboob, F.; Badshah, N.; Jabeen, N.; Ayub, G.

2004-01-01

Research work was conducted to find out the effect of irradiation on the post-harvest life of potatoes. Cultivar Raja was obtained from Agricultural Research Institute, Tarnab, and irradiated by Cobalt-60 source at different doses 0, 5, 7.5, 10 and 15 Krad at the Nuclear Institute for Food and Agriculture (NIFA), Tarnab during the year 2002. The samples were then stored for three months at the Horticultural Research Farm, Malakandher, at a room temperature of 30-39 degree C and relative humidity of 29-63%. Various tests carried out at Food Science laboratory revealed that irradiation significantly affected the weight loss, sugars, starch, ascorbic acid, sprouting and specific gravity. It was observed that maximum sprouting has occurred in control (42.1%) followed by 5 Krad irradiated tubers (6.4%). While irradiation doses of 7.5, 10 and 15 Krad completely inhibited sprouting. Maximum percent decrease in weight (42.66%), reducing sugars (0.57%), non reducing sugars (0.87%), starch (12%), ascorbic acid (32%) and specific gravity (4%) were recorded for control while minimum percent decrease in weight (31.40%), reducing sugars (0.19%), non-reducing sugars (0.27%), starch (8.0%), ascorbic acid (12%) and specific gravity (1.7%) were noted for IS Krad irradiated tubers. Irradiation dose of 7.5 Krad seems to be better for the extension of shelf life of potatoes

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.; Robert D. Mariani; Rory Kennedy; Doug Toomer

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States’ ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

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)

Post-Impact and Open Hole Tensile Of Kenaf Hybrid Composites

Science.gov (United States)

Yunus, S.; Salleh, Z.; Masdek, N. R. N. M.; Taib, Y. M.; Azhar, I. I. S.; Hyie, K. M.

2018-03-01

Nowadays, kenaf hybrid glass composites has been used for a vast field of study throughout the globe. There are several compositions and orientation of kenaf hybrid glass composites that has been studied. With regards to the study that has been done, this study will be focussing on a 90FG/0/90/90/0/90FG orientation of kenaf hybrid glass composites. Polyester resin is used as a matrix to these hybrid composites. Impacted and open hole specimens were then analyzed through tensile test. All specimens were fabricated by using the cold press hand lay-up technique. The results revealed that the hybrid composites were hardly affected by the impact up to 6J. After 6J the impacted specimens experienced a significant damage for both strength and modulus. The same goes to open hole specimens where the same trend of tensile properties were observed as impacted specimens.

Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

2000-03-01

Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Alexandre - a multi-project, multi-material and multi-technique action for an irradiation experiment in Osiris and post irradiation examination

International Nuclear Information System (INIS)

Averty, X.; Brachet, J.C.; Bertin, J.L.; Pizzanelli, J.P.; Rozenblum, F.

1999-01-01

This paper presents the data obtained on different classes of steels neutron irradiated at 325 deg C in pressurized water with a PWR-type chemistry. This irradiation, nicknamed 'Alexandre', took place in the OSIRIS reactor and finished in November 1999, for a maximum irradiation damage of ∼9 dpa. The preliminary results (up to 3.4 dpa), discussed in relation to chemical composition and initial metallurgical conditions, are listed below: - Evolution of the mechanical properties as a function of irradiation dose including the measurements of the Reduction-in-Area to failure by image analysis. - Comparison between out-of-pile and in-pile uniform corrosion. - Microstructural aspects (fractography, Transmission Electron Microscopy, and Small Angle Neutron Scattering measurements). - Post-irradiation evolution of residual. activity. (authors)

Post-irradiation dietary vitamin E does not affect the development of radiation-induced lung damage in rats

International Nuclear Information System (INIS)

Wiegman, Erwin M.; Gameren, Mieke M. van; Kampinga, Harm H.; Szabo, Ben G.; Coppes, Rob P.

2004-01-01

The purpose of this study was to investigate whether application of post-irradiation vitamin E, an anti-oxidant, could prevent the development of radiation induced lung damage. Wistar rats were given vitamin E enriched or vitamin E deprived food starting from 4 weeks after 18 Gy single dose irradiation of the right thorax. Neither breathing frequencies nor CT density measurements revealed differences between the groups. It is concluded that post-irradiation vitamin E does not influence radiation-induced fibrosis to the lung

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)

The tensile strength of mechanical joint prototype of lontar fiber composite

Science.gov (United States)

Bale, Jefri; Adoe, Dominggus G. H.; Boimau, Kristomus; Sakera, Thomas

2018-03-01

In the present study, an experimental activity has been programmed to investigate the effect of joint prototype configuration on tensile strength of lontar (Borassus Flabellifer) fiber composite. To do so, a series of tests were conducted to establish the tensile strength of different joint prototype configuration specimen of lontar fiber composite. In addition, post observation of macroscope was used to map damage behavior. The analysis of lontar fiber composite is a challenge since the material has limited information than others natural fiber composites materials. The results shown that, under static tensile loading, the tensile strength of 13 MPa produced by single lap joint of lontar fiber composite is highest compare to 11 MPa of tensile strength generated by step lap joint and double lap joint where produced the lowest tensile strength of 6 MPa. It is concluded that the differences of tensile strength depend on the geometric dimensions of the cross-sectional area and stress distribution of each joint prototype configuration.

AECL hot-cell facilities and post-irradiation examination services

International Nuclear Information System (INIS)

Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

1998-04-01

This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analyses are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

AECL hot-cell facilities and post-irradiation examination services

International Nuclear Information System (INIS)

Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

1995-01-01

This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analysis are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

Effects of heat treatments and neutron irradiation on the physical and mechanical properties of copper alloys at 100 deg. C

International Nuclear Information System (INIS)

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

1998-05-01

The final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys is described herein. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. Additional specimens were reaged and given a reactor bakeout treatment at 350 deg. C for 100 h. GlidCop TM CuAl-15 (previously referred to as CuAl-25) was given a heat treatment corresponding to a bonding thermal cycle only. Specimens were neutron irradiated at 100 deg. C to a dose level of ∼0.3 dpa. Post-irradiation tensile tests at (100 deg. C), electrical resistivity measurements (at 23 deg. C), and microstructural examinations were performed. The post-irradiation tests at 100 deg. C revealed that the greatest loss of ductility occurred in the CuCrZr alloys irradiated at 100 deg. C, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which exhibited much higher uniform elongation and strength after irradiation than that observed in the case of CuCrZr alloys. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The CuAl-25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure

Post-{gamma}-irradiation reactions in vitamin E stabilised and unstabilised HDPE

Energy Technology Data Exchange (ETDEWEB)

Mallegol, J.; Carlsson, D.J. E-mail: dave.carlsson@nrc.ca; Deschenes, L

2001-12-01

The oxidation of high density polyethylene (HDPE), both unstabilised and vitamin E stabilised, has been studied by infrared (IR) and electron paramagnetic resonance (EPR) spectroscopies in the period following {gamma}-irradiation at doses from 1 to 60 kGy (range of food sterilisation). Derivatisation by reaction with sulphur tetrafluoride was used to identify macro-ketone and carboxylic acid components of the overlapped IR carbonyl region. Oxidation continued for several hundred hours after the cessation of irradiation as shown by the increase in hydroxyl, ketone and acid groups. Carboxylic acid groups are particularly important as a direct indication of backbone scission. Vitamin E, although an effective antioxidant during {gamma}-irradiation, was less effective in reducing the post-irradiation changes, which are probably driven by migration of radical sites along the polymer backbone from within the crystalline phase to the amorphous/crystalline inter-phase, where they become oxygen accessible.

The exercise-induced biochemical milieu enhances collagen content and tensile strength of engineered ligaments.

Science.gov (United States)

West, Daniel W D; Lee-Barthel, Ann; McIntyre, Todd; Shamim, Baubak; Lee, Cassandra A; Baar, Keith

2015-10-15

Exercise stimulates a dramatic change in the concentration of circulating hormones, such as growth hormone (GH), but the biological functions of this response are unclear. Pharmacological GH administration stimulates collagen synthesis; however, whether the post-exercise systemic milieu has a similar action is unknown. We aimed to determine whether the collagen content and tensile strength of tissue-engineered ligaments is enhanced by serum obtained post-exercise. Primary cells from a human anterior cruciate ligament (ACL) were used to engineer ligament constructs in vitro. Blood obtained from 12 healthy young men 15 min after resistance exercise contained GH concentrations that were ∼7-fold greater than resting serum (P Ligament constructs were treated for 7 days with medium supplemented with serum obtained at rest (RestTx) or 15 min post-exercise (ExTx), before tensile testing and collagen content analysis. Compared with RestTx, ExTx enhanced collagen content (+19%; 181 ± 33 vs. 215 ± 40 μg per construct P = 0.001) and ligament mechanical properties - maximal tensile load (+17%, P = 0.03 vs. RestTx) and ultimate tensile strength (+10%, P = 0.15 vs. RestTx). In a separate set of engineered ligaments, recombinant IGF-1, but not GH, enhanced collagen content and mechanics. Bioassays in 2D culture revealed that acute treatment with post-exercise serum activated mTORC1 and ERK1/2. In conclusion, the post-exercise biochemical milieu, but not recombinant GH, enhances collagen content and tensile strength of engineered ligaments, in association with mTORC1 and ERK1/2 activation. © 2015 The Authors. The Journal of Physiology © 2015 The Physiological Society.

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

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.)

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.

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.

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.)

Post irradiation effects on the graft of poly(tetrafluoroethylene-co-perfluoropropyl vinyl ether) (PFA) films

International Nuclear Information System (INIS)

Geraldes, Adriana N.; Zen, Heloisa A.; Ribeiro, Geise; Ferreira, Henrique P.; Souza, Camila P.; Parra, Duclerc F.; Lugao, Ademar B.

2009-01-01

Radiation induced grafting of monomers into fluorinated polymers was designed as an alternative route to polymer modification. In this work, grafting of styrene onto poly(tetrafluoroethylene-co-perfluoropropyl vinyl ether) (PFA) was studied. Radiation-induced grafting of styrene onto PFA films was investigated after simultaneous irradiation (in post-irradiation condition) using a 60 Co source. The films of PFA were irradiated at 20, 40, 80 and 100 kGy doses at room temperature and chemical changes were monitored after contact with styrene for grafting. The post-irradiation time was established between 7 and 28 days when films of PFA were maintained in styrene/toluene 1:1 v/v solution at room temperature. After these periods the grafting degrees were evaluated in the samples. The highest degree of grafting was achieved after 14 days. Chemical modifications were evaluated by infrared spectroscopic analysis (FTIR), thermogravimetry (TG), differential scanning calorimetry (DSC) and also by scanning electron microscopy (SEM). The degree of grafting (DOG) was determined gravimetrically. The results showed that irradiated PFA films at 100 kGy exhibited higher grafting degree. Surface analysis by SEM technique of irradiated, grafted and original films have presented an homogeneous surface. (author)

On the determination of the post-irradiation time from the glow curve of TLD-100

International Nuclear Information System (INIS)

Weinstein, M.; German, U.; Dubinsky, S.; Alfassi, Z.B.

2003-01-01

The ratio of peak 3 to the sum of peaks 4 + 5 in TLD-100 was measured for various pre-irradiation and post-irradiation time periods, under conditions characteristic of routine personal dosimetry. It was confirmed that the value of this ratio depends only on the elapsed time between the prior readout and the present one, independent of the moment when the irradiation took place during the total time interval (storage time). This effect indicates that fading of peak 3 seems to be due mainly to changes in the unoccupied traps, and not to decay of trapped charges, being almost independent of the presence of electrons or holes in the traps. This observation leads to the conclusion that the suggestions in the past to use the decay of peak 3 in TLD-100 for the measurement of the elapsed time between irradiation and readout may have been wrong. On the other hand, the decay of peak 2 can be used to measure the elapsed time from irradiation, since the rate of decay is different when related to pre-irradiation and post-irradiation times, indicating a much higher decay rate of the trapped charges (Randall-Wilkins decay). However, because of the fast decay rate of peak 2, its use for determination of the elapsed time since irradiation is of little practical significance. (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

Development status of post irradiation examination techniques at the JMTR Hot Laboratory

International Nuclear Information System (INIS)

Ohmi, M.; Ohsawa, K.; Nakagawa, T.; Umino, A.; Shimizu, M.; Satoh, H.; Oyamada, R.

1992-01-01

Hot laboratory at Oarai Research Establishment was founded to examine the objects mainly irradiated at JMTR (Japan Materials Testing Reactor) and has been operated since 1971. A wide variety of post-irradiation examinations (PIE) is available using the hot laboratory. Continuous efforts are made to develop new PIE techniques to accommodate the user's requirements. The following are main techniques recently developed in the hot laboratory; 1. Remote capsule assembly including remote weld of irradiated objects for reirradiation in JMTR. 2. Fracture toughness tests of reactor component materials. 3. Creep tests of heat resistance alloys in high temperature conditions. 4. Tests of irradiation assisted stress corrosion cracking (IASCC). 5. Examination techniques of miniaturized test specimens. This report describes an outline of the hot laboratory with main emphasis on the new PIE techniques. (author)

Post-irradiation examination of a 13000C-HTR fuel experiment Project J 96.M3

International Nuclear Information System (INIS)

Bueger, J. de; Roettger, H.

1977-01-01

A large variety of loose coated fuel particles have been irradiated in the BR2 at Mol/Belgium at temperatures between 1200 0 C and 1400 0 C and up to a fast neutron fluence of 1.2x1022 cm -2 (E>0.1 MeV) as a Euratom sponsored experiment for the advanced testing of HTR fuel. The specimens have been provided by Belgonucleaire and the Dragon Project. A short description of the experiment as well as the results of post-irradiation examination mainly carried out at Petten (N.H.), The Netherlands, are presented here. The post-irradiation examination has shown that the required performance can be achieved by a number of the tested fuel specimens without serious damage

Kinetic Monte Carlo simulation of nanostructural evolution under post-irradiation annealing in dilute FeMnNi

Energy Technology Data Exchange (ETDEWEB)

Chiapetto, M. [SCK-CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, C.S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Laboratoire commun EDF-CNRS, Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Domain, C. [EDF R and D, Departement Materiaux et Mecanique des Composants, Les Renardieres, Moret sur Loing (France); Laboratoire commun EDF-CNRS, Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Mol (Belgium)

2015-01-01

Post-irradiation annealing experiments are often used to obtain clearer information on the nature of defects produced by irradiation. However, their interpretation is not always straightforward without the support of physical models. We apply here a physically-based set of parameters for object kinetic Monte Carlo (OKMC) simulations of the nanostructural evolution of FeMnNi alloys under irradiation to the simulation of their post-irradiation isochronal annealing, from 290 to 600 C. The model adopts a ''grey alloy'' scheme, i.e. the solute atoms are not introduced explicitly, only their effect on the properties of point-defect clusters is. Namely, it is assumed that both vacancy and SIA clusters are significantly slowed down by the solutes. The slowing down increases with size until the clusters become immobile. Specifically, the slowing down of SIA clusters by Mn and Ni can be justified in terms of the interaction between these atoms and crowdions in Fe. The results of the model compare quantitatively well with post-irradiation isochronal annealing experimental data, providing clear insight into the mechanisms that determine the disappearance or re-arrangement of defects as functions of annealing time and temperature. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

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

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)

Post-irradiation examination of Oconee 1 fuel - cycle 1 destructive test phase

International Nuclear Information System (INIS)

1979-07-01

Standard B and W Mark-B (15 x 15) pressurized water reactor fuel rods were destructively examined after one cycle of irradiation in the Oconee 1 reactor. Fuel rod average burnup ranged from 10,603 to 11,270 MWd/mtU for the rods examined. Data obtained included fuel rod extraction loads, rod dimensional changes, cladding tensile properties, fuel pellet gap length, fission product distribution, fission gas and crud composition, fuel densification, chemical burnup analysis, and fuel and cladding microstructure. As expected, parametric changes were well within the design envelope. Superficial corrosion and wear were found at spacer grid contact points. However, the 19 rods examined were structurally sound and exhibited no indications of cladding defects associated with pelletcladding interactions

Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

International Nuclear Information System (INIS)

Freund, D.; Elbel, H.; Steiner, H.

1976-06-01

The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

RERTR-12 Post-irradiation Examination Summary Report

Energy Technology Data Exchange (ETDEWEB)

Rice, Francine [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Walter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robinson, Adam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabin, Barry [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

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.

Effect of ITER components manufacturing cycle on the irradiation behaviour of 316L(N)-IG steel

International Nuclear Information System (INIS)

Rodchenkov, B.S.; Prokhorov, V.I.; Makarov, O.Yu.; Shamardin, V.K.; Kalinin, G.M.; Strebkov, Yu.S.; Golosov, O.A.

2000-01-01

The main options for the manufacturing of high heat flux (HHF) components is hot isostatic pressing (HIP) using either solid pieces or powder. There was no database on the radiation behaviour of these materials, and in particular stainless steel (SS) 316L(N)-IG with ITER components manufacturing thermal cycle. Irradiation of wrought steel, powder-HIP, solid-HIP and HIPed joints has been performed within the framework of an ITER task. Specimens cut from 316L(N)-IG plate, HIP products, and solid-HIP joints were irradiated in the SM-3 reactor in Dimitrovgrad up to 4 and 10 dpa at 175 deg. C and 265 deg. C. The paper describes the results of post-irradiation tensile and fracture toughness tests

Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

International Nuclear Information System (INIS)

Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.

2001-05-01

The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project T he Nuclear Fuel Material Development of Research Reactor . And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

Post-irradiation treatment of human lymphocytes with spermidine reduced frequency of chromatid breaks

International Nuclear Information System (INIS)

Bocian, E.; Rosiek, O.; Ziemba-Zoltowska, B.

1978-01-01

Human lymphocyte cultures were X-irradiated with a single dose of 100 or 200 rad 46 h after phytohemagglutinin stimulation. In dose-fractionation experiments, 2h later the second dose was applied. All the cultures were harvested at 54 h after their initiation. In lymphocytes irradiated with a single dose of 200 rad, 2h post-irradiation contact with 10 -5 M exogeneous spermidine resulted in reduction of chromatid breaks by 34 %. Introduction of spermidine into culture medium for fractionation interval between the 2 doses of 100 rad reduced the frequency of chromatid breaks by 42 %. (author)

Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

International Nuclear Information System (INIS)

Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

2013-06-01

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

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 ...

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.)

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

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)

Effect of Irradiation Maternal Diets on the Post-natal Development of Brain Rat Pups

International Nuclear Information System (INIS)

Hasan, S.S.

2005-09-01

Full text: Effect of Protein-calorie malnutrition was studied on the pups born to mothers receiving either irradiated normal diet (consisted equal parts of gram and wheat) or irradiation low protein diet (consisted one part of normal diet and three parts of heat). Level of DNA, RNA and protein content were found markedly reduced in the brain of irradiated low protein diet fed pups than in the pups fed on the irradiated normal diet. Glucose 6-phosphate dehydrogenase activity was found lower while catalase and lipid peroxidation activity were higher in the pups given irradiated low protein diet, compared whit the pups fed irradiated normal diet. On the whole both the irradiated low protein diet as well as irradiated normal diet fed pups showed higher index of biochemical changes than in the unirradiated low protein diet fed pups. Post-natal mortality was 60% in the pups given irradiated low protein diet, whereas the pups fed on the irradiated normal diet and unirradiated low protein diet did not show any death. The study given evidence that feeding of the irradiated low protein diet interferes more with the development of brain compared with the pups fed on irradiated normal diet

Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-01-01

This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U 3 Si fuel irradiated in the NRU reactor. (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

Gamma radiolysis and post-irradiation leaching of ion exchange resins

International Nuclear Information System (INIS)

Traboulsi, A.

2012-01-01

The knowledge of the behavior under irradiation and in presence of water of Ion Exchange Resins (IER) is very necessary to predict their impact on the environment during the storage phase and in a possible deep geological disposal. The IER studied are the MB400 mixed bed resin and its 'pure' anionic and cationic components. The experimental strategy used in this work was based on the use of chemometric tools permitting to estimate the effect of the irradiation atmosphere, the dose rate, the absorbed dose and the leaching temperature. The gaseous and water-soluble radiolysis products were analyzed by gas Mass Spectrometry (MS) and Ion Chromatography (IC). The IER generated principally H 2 g, CO 2 g and amines for which quantities depended of the resin nature and the irradiation conditions. The analysis of solid irradiated resins was investigated by Fourier Transformed Infrared (FTIR) and Nuclear Magnetic Resonance ( 13 C NMR) techniques. The last ones revealed structural modifications of the IER solid matrix in function of the experimental conditions. Their behavior in presence of water was studied during 143 days by characterization of the organic matter released after their post-irradiation leaching. The kinetics showed that all the water-soluble components were releasing at the first contact with water. The Total Organic Carbon (TOC) quantity released depends, according to the resin nature, either on the dose, either on the irradiation atmosphere. The dose rate has no effect on the degradation and the leaching of the MB400 resin, which behaved differently than its pure components. (author) [fr

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)

Solvent effect on post-irradiation grafting of styrene onto poly(ethylene-alt-tetrafluoroethylene) (ETFE) films

Science.gov (United States)

Napoleão Geraldes, Adriana; Augusto Zen, Heloísa; Ribeiro, Geise; Fernandes Parra, Duclerc; Benévolo Lugão, Ademar

2013-03-01

Radiation-induced grafting of styrene onto ETFE films in different solvent was investigated after simultaneous irradiation (in post-irradiation condition) using a 60Co source. Grafting of styrene followed by sulfonation onto poly(ethylene-alt-tetrafluoroethylene) (ETFE) are currently studied for synthesis of ion exchange membranes. The ETFE films were immersed in styrene/toluene, styrene/methanol and styrene/isopropyl alcohol and irradiated at 20 and 100 kGy doses at room temperature. The post-irradiation time was established at 14 day and the grafting degree was evaluated. The grafted films were sulfonated using chlorosulfonic acid and 1,2-dichloroethane 20:80 (v/v) at room temperature for 5 h. The degree of grafting (DOG) was determined gravimetrically and physical or chemical changes were evaluated by differential scanning calorimeter analysis (DSC), thermogravimetric analysis (TGA) and scanning electron microscopy (SEM). The ion exchange capacity (IEC) values showed the best performance of sulfonation for ETFE membranes grafted in toluene solvent. Surface images of the grafted films by SEM technique have presented a strong effect of the solvents on the films morphology.

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)

Gamma-ray irradiation and post-irradiation at room and elevated temperature response of pMOS dosimeters with thick gate oxides

Directory of Open Access Journals (Sweden)

Pejović Momčilo M.

2011-01-01

Full Text Available Gamma-ray irradiation and post-irradiation response at room and elevated temperature have been studied for radiation sensitive pMOS transistors with gate oxide thickness of 100 and 400 nm, respectively. Their response was followed based on the changes in the threshold voltage shift which was estimated on the basis of transfer characteristics in saturation. The presence of radiation-induced fixed oxide traps and switching traps - which lead to a change in the threshold voltage - was estimated from the sub-threshold I-V curves, using the midgap technique. It was shown that fixed oxide traps have a dominant influence on the change in the threshold voltage shift during gamma-ray irradiation and annealing.

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)

Post-irradiation pericardial malignant mesothelioma with deletion of p16: a case report.

Science.gov (United States)

Naeini, Yalda B; Arcega, Ramir; Hirschowitz, Sharon; Rao, Nagesh; Xu, Haodong

2018-02-01

Malignant mesotheliomas are rather uncommon neoplasms associated primarily with asbestos exposure; however, they may also arise as second primary malignancies after radiation therapy, with a latency period of 15-25 years. Numerous studies have reported an association between pleural malignant mesothelioma and chest radiation performed for other malignancies; on the other hand, post-irradiation mesotheliomas of the pericardium have been reported in only a few published cases to date, and no homozygous deletion of 9p21 has been described in such cases. We report the case of a 48-year-old man with a history of Hodgkin's lymphoma and no prior asbestos exposure who developed pericardial malignant epithelioid mesothelioma. We further discuss the cytologic, histologic, immunophenotypic, and fluorescence in situ hybridization findings in this case. To our knowledge, this is the first well-documented case of post-radiation pericardial malignant mesothelioma showing homozygous deletion of 9p21. Homozygous deletion of 9p21, the locus harboring the p16 gene, is present in post-irradiation pericardial malignant mesothelioma.

Effect of the hydrolytic state of dietary protein on post-irradiation morbidity and mucosal cell regeneration

International Nuclear Information System (INIS)

Beitler, M.K.; Mahler, P.A.; Yamanaka, W.K.; Guy, D.G.; Hutchinson, M.L.

1987-01-01

Diets containing hydrolyzed casein have been observed to enhance post-irradiation intestinal mucosal recovery. The intake and the composition of such diets were not carefully controlled. This study attempted to do so. Male specific pathogen-free Sprague-Dawley rats were randomized to receive either an enzymatically hydrolyzed casein semi-purified diet (EHC), a whole casein semi-purified diet (WC), or powdered lab chow (C). All diets were isonitrogenous, and the WC and C rats were pair-fed to the ad libitum fed EHC rats. Seven days after initiation of feeding, the rats were abdominally irradiated with a single 9.0 Gy dose of 137Cs gamma rays. The rats were continued on the diets for another 5 days. Intestinal mucosa from transverse segments at the duodenum, jejunum, proximal ileum, and distal ileum were measured for incorporation of ( 3 H methyl) thymidine 1 hour after intraperitoneal injection. Incorporation reached a maximum by day 4 post-irradiation regardless of diet or segment. Incorporation in the duodenum was enhanced by the EHC diet compared to the C diet, while the incorporation in the jejunum was initially suppressed by the EHC diet compared to the WC diet. In the jejunum, the number of mitoses per crypt of 25 anti-mesenteric crypts post-irradiation was increased by the EHC diet. Prior to irradiation, all groups gained similar amounts of weight. After irradiation, the C rats lost weight, while the EHC and WC rats remained the same or gained weight. Guaiac tests for occult blood were negative prior to irradiation, but positive for all rats on days 1-5 postirradiation. When calorie and protein intakes were controlled, different areas of the small intestine responded differently to EHC

Post-harvest UVC irradiation effect on anthocyanin profile of grape berries

International Nuclear Information System (INIS)

Rosas, I. de; Ponce, M.; Gargantini, R.; Martinez, L.

2010-01-01

Anthocyanins are a class of phenolic compounds that contribute to the color of red grapes and have shown nutraceutical properties for human health. UVC light irradiation has been proved to increase phenolic compounds such as stilbenes, but its effect on anthocyanins has not been reported. The aim of this work was to identify the best treatment conditions of UVC light irradiation on post-harvest berries of Malbec (M), Cabernet Sauvignon (CS) and Tempranillo (T) for anthocyanin increments. Grape berries were irradiated with 240 W at 20 and 40 cm from the light source, for 30, 60 and 120 seconds. Both, irradiated and control grapes were stored on darkness at 20 C degree until anthocyanin extraction with methanol/ClH. HPLC analysis were performed and nine anthocyanins were quantified. UVC light irradiation modified the anthocyanin profile of the three cultivars. All the glucoside anthocyanins derivates and peonidin-acetyl-glucoside, as well as total anthocyanins were increased when CS berries were exposed to UVC for 120 s at 40 cm. This suggests that UVC stimulated the entire biosynthetic pathway. The anthocyanin content of the control berries was always higher than the treatments with UVC on M and T, making necessary to evaluate less rigorous conditions for these varieties. (authors)

Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

International Nuclear Information System (INIS)

Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

2011-01-01

The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

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)

Post-curing conversion kinetics as functions of the irradiation time and increment thickness

Directory of Open Access Journals (Sweden)

Nicola Scotti

2013-04-01

Full Text Available Objective: This study evaluated the variation of conversion degree (DC in the 12 hours following initial photoactivation of a low-shrinkage composite resin (Venus Diamond. Material and Methods: The conversion degree was monitored for 12 hours using Attenuated Total Reflection (ATR F-TIR Spectroscopy. The composite was placed in 1 or 2 mm rings and cured for 10 or 20 seconds with a LED lamp. ATR spectra were acquired from the bottom surface of each sample immediately after the initial photoactivation (P=0, 30 minutes (P=0.5 and 12 hours after photoactivation (P=12 in order to obtain the DC progression during the post-curing period. Interactions between thickness (T, irradiation time (I and post-curing (P on the DC were calculated through ANOVA testing. Results: All the first order interactions were statistically significant, with the exception of the T-P interaction. Furthermore, the shift from P=0 to P=0.5 had a statistically higher influence than the shift from P=0.5 to P=12. The post-curing period played a fundamental role in reaching higher DC values with the low-shrinkage composite resin tested in this study. Moreover, both the irradiation time and the composite thickness strongly influenced the DC. Conclusions: Increased irradiation time may be useful in obtaining a high conversion degree (DC with a low-shrinkage nano-hybrid composite resin, particularly with 2 mm composite layers.

Microstructural evolution of nanochannel CrN films under ion irradiation at elevated temperature and post-irradiation annealing

Science.gov (United States)

Tang, Jun; Hong, Mengqing; Wang, Yongqiang; Qin, Wenjing; Ren, Feng; Dong, Lan; Wang, Hui; Hu, Lulu; Cai, Guangxu; Jiang, Changzhong

2018-03-01

High-performance radiation tolerance materials are crucial for the success of future advanced nuclear reactors. In this paper, we present a further investigation that the "vein-like" nanochannel films can enhance radiation tolerance under ion irradiation at high temperature and post-irradiation annealing. The chromium nitride (CrN) nanochannel films with different nanochannel densities and the compact CrN film are chosen as a model system for these studies. Microstructural evolution of these films were investigated using Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Elastic Recoil Detection (ERD) and Grazing Incidence X-ray Diffraction (GIXRD). Under the high fluence He+ ion irradiation at 500 °C, small He bubbles with low bubble densities are observed in the irradiated nanochannel CrN films, while the aligned large He bubbles, blistering and texture reconstruction are found in the irradiated compact CrN film. For the heavy Ar2+ ion irradiation at 500 °C, the microstructure of the nanochannel CrN RT film is more stable than that of the compact CrN film due to the effective releasing of defects via the nanochannel structure. Under the He+ ion irradiation and subsequent annealing, compared with the compact film, the nanochannel films have excellent performance for the suppression of He bubble growth and possess the strong microstructural stability. Basing on the analysis on the sizes and number densities of bubbles as well as the concentrations of He retained in the nanochannel CrN films and the compact CrN film under different experimental conditions, potential mechanism for the enhanced radiation tolerance are discussed. Nanochannels play a crucial role on the release of He/defects under ion irradiation. We conclude that the tailored "vein-like" nanochannel structure may be used as advanced radiation tolerance materials for future nuclear reactors.

Post irradiation conical keratosis

International Nuclear Information System (INIS)

Vestey, J.P.; Hunter, J.A.A.; Mallet, R.B.; Rodger, A.

1989-01-01

The authors have recently seen 3 patients affected by a widespread eruption of minute keratoses confined to areas of irradiated skin with clinical and histologial features of which they have been unable to find previous literary descriptions. A fourth patient with similar clinical and histopathological features occurring after exposure only to actinic irradiation is described. (author)

Post irradiation conical keratosis

Energy Technology Data Exchange (ETDEWEB)

Vestey, J.P.; Hunter, J.A.A. (Royal Infirmary, Edinburgh (UK)); Mallet, R.B. (Westminster Hospital, London (UK)); Rodger, A. (Western General Hospital, Edinburgh (UK))

1989-03-01

The authors have recently seen 3 patients affected by a widespread eruption of minute keratoses confined to areas of irradiated skin with clinical and histologial features of which they have been unable to find previous literary descriptions. A fourth patient with similar clinical and histopathological features occurring after exposure only to actinic irradiation is described. (author).

Influence of UV Photo-Transfer on Post Irradiated Double Sulphate Poly-Crystals By Gamma And X-rays

International Nuclear Information System (INIS)

El-kolaly, M.A.

2000-01-01

Solid state thermoluminescence (TL) dosimetry has for many years been the pre-eminent method for quantifying ionizing radiation dose. In this work, thermoluminescence characteristics of the double sulphate (Li Cs So 4 ) poly-crystals have been studied after exposure to different doses from X and gamma radiation. The glue curves showed TL response of three peaks at 75,125,250 degree. The structure of the glue peaks due to X-rays is quite different from that due to gamma rays. UV exposure yields a regeneration of the TL peaks for the post irradiated samples for X or gamma radiation with some changes in the peaks structure especially the third peak. For the post X-ray irradiated crystals, the area under the third glow peak (III) increased linearly with the integrated time of UV exposures till about 30 min. after which no changes were observed; while , for the post gamma-irradiated crystals two linear regions were observed

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

Post irradiation examination of control rod assembly of FBTR

International Nuclear Information System (INIS)

Anandaraj, V.; Raghu, N.; Venkiteswaran, C.N.; Visweswaran, P.; Vijayakumar, Ran; Jayaraj, V.V.; Padmaprabu, P.; Saravanan, T.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.

2010-01-01

Six control rods with boron carbide pellets are used in FBTR for shutdown and control of reactor power. One control rod after being subjected to a fluence level of 7.2 x 10 22 n/cm 2 was received for post irradiation examination (PIE) to assess its irradiation behavior and to investigate the incident of dropping of control rod. Examinations carried out include precise dimensional measurements to investigate the possibility of interference between the control rod and outer sheath, Neutron radiography and x-radiograph to assess the integrity of the boron carbide pellets and other internals, density measurements to assess the swelling behaviour of boron carbide pellets and metallographic examinations to study the cracking behaviour and microstructural changes in the pellet and the clad. Depletion of B 10 in the pellet was studied using time of flight mass spectrometry. The paper highlights the examinations and results of the PIE carried out. (author)

Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-09-01

This paper describes the post-irradiation examination of 4 intact low-enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 0 coolant inlet temperature 37 degrees C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

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)

Effect of post weld heat treatment on the microstructure and tensile properties of activated flux TIG welds of Inconel X750

Energy Technology Data Exchange (ETDEWEB)

Ramkumar, K. Devendranath, E-mail: ramdevendranath@gmail.com; Ramanand, R.; Ameer, Ajmal; Simon, K. Aghil; Arivazhagan, N.

2016-03-21

This study addresses the effect of post weld heat treatment on the fusion zone microstructure and the mechanical properties of activated flux tungsten inert gas (A-TIG) weldments of Inconel X750. In this study, a compound flux of 50% SiO{sub 2}+50% MoO{sub 3} was used for A-TIG welding of the samples. Comparative studies on the microstructure and mechanical properties have been made on the weldments both in the as-welded and post weld heat treated conditions. Direct ageing post weld heat treatment (PWHT) was carried out at 705 °C for 22 h on the A-TIG weldment to assess the structure–property relationships. It was inferred that direct ageing post weld heat treatment resulted in better tensile strength (1142 MPa) compared to the as-welded coupons (736 MPa). The joint efficiencies of the as-welded and post weld heat treated conditions were found to be 60.7% and 94.07% respectively. The impact toughness of the as-welded coupons were found to be greater than the post weld heat treated samples; however the impact toughness of the welds are greater than the parent metal employed in both the cases. This study also attested the detailed structure–property relationships of A-TIG weldments using the combined techniques of optical and scanning electron microscopy, Electron Dispersive X-ray Analysis (EDAX) techniques.

Effect of post weld heat treatment on the microstructure and tensile properties of activated flux TIG welds of Inconel X750

International Nuclear Information System (INIS)

Ramkumar, K. Devendranath; Ramanand, R.; Ameer, Ajmal; Simon, K. Aghil; Arivazhagan, N.

2016-01-01

This study addresses the effect of post weld heat treatment on the fusion zone microstructure and the mechanical properties of activated flux tungsten inert gas (A-TIG) weldments of Inconel X750. In this study, a compound flux of 50% SiO_2+50% MoO_3 was used for A-TIG welding of the samples. Comparative studies on the microstructure and mechanical properties have been made on the weldments both in the as-welded and post weld heat treated conditions. Direct ageing post weld heat treatment (PWHT) was carried out at 705 °C for 22 h on the A-TIG weldment to assess the structure–property relationships. It was inferred that direct ageing post weld heat treatment resulted in better tensile strength (1142 MPa) compared to the as-welded coupons (736 MPa). The joint efficiencies of the as-welded and post weld heat treated conditions were found to be 60.7% and 94.07% respectively. The impact toughness of the as-welded coupons were found to be greater than the post weld heat treated samples; however the impact toughness of the welds are greater than the parent metal employed in both the cases. This study also attested the detailed structure–property relationships of A-TIG weldments using the combined techniques of optical and scanning electron microscopy, Electron Dispersive X-ray Analysis (EDAX) techniques.

Effect of UV on the post irradiated Li Cs S O{sub 4} crystal by X and gamma radiation. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Kassem, M E [On Leave, Alexandria University, Faculty of Science, PHysics Department. Alexandria (Egypt); EL-Kolaly, M A [On Leave, Radiation Protection Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Al-Houty, L I [University of Qatar, Faculty of Science, Department of Physics, P.O. Box 2713, Doha (Qatar)

1996-03-01

Thermoluminescence characteristics of Li Cs S O{sub 4} crystal have been studied after exposure to different doses of X and Gamma-Radiations. The glow curves showed TL response of Three peaks at 75,125 and 250 degree C. The structure of the glow peaks due to X-rays is quite different from that due to gamma-rays. UV exposure yields regeneration of the TL peaks for the post irradiated samples with X or Gamma-radiation with some changes in the peaks`s structure especially the third peak. For the post X-ray irradiated crystals, the area under the third glow peak (PK III) increased with integrated time of UV exposure till about 30 min after which no changes were observed; while, for the post gamma-irradiated crystals, two linear regions were observed. The models of the TL response for the post irradiated samples as a result to UV are discussed. 5 figs.

A stochastic post-processing method for solar irradiance forecasts derived from NWPs models

Science.gov (United States)

Lara-Fanego, V.; Pozo-Vazquez, D.; Ruiz-Arias, J. A.; Santos-Alamillos, F. J.; Tovar-Pescador, J.

2010-09-01

Solar irradiance forecast is an important area of research for the future of the solar-based renewable energy systems. Numerical Weather Prediction models (NWPs) have proved to be a valuable tool for solar irradiance forecasting with lead time up to a few days. Nevertheless, these models show low skill in forecasting the solar irradiance under cloudy conditions. Additionally, climatic (averaged over seasons) aerosol loading are usually considered in these models, leading to considerable errors for the Direct Normal Irradiance (DNI) forecasts during high aerosols load conditions. In this work we propose a post-processing method for the Global Irradiance (GHI) and DNI forecasts derived from NWPs. Particularly, the methods is based on the use of Autoregressive Moving Average with External Explanatory Variables (ARMAX) stochastic models. These models are applied to the residuals of the NWPs forecasts and uses as external variables the measured cloud fraction and aerosol loading of the day previous to the forecast. The method is evaluated for a set one-moth length three-days-ahead forecast of the GHI and DNI, obtained based on the WRF mesoscale atmospheric model, for several locations in Andalusia (Southern Spain). The Cloud fraction is derived from MSG satellite estimates and the aerosol loading from the MODIS platform estimates. Both sources of information are readily available at the time of the forecast. Results showed a considerable improvement of the forecasting skill of the WRF model using the proposed post-processing method. Particularly, relative improvement (in terms of the RMSE) for the DNI during summer is about 20%. A similar value is obtained for the GHI during the winter.

Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

International Nuclear Information System (INIS)

Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

2005-01-01

Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

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)

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.

Strain distribution during tensile deformation of nanostructured aluminum samples

DEFF Research Database (Denmark)

Kidmose, Jacob; Lu, L.; Winther, Grethe

2012-01-01

To optimize the mechanical properties, especially formability, post-process deformation by cold rolling in the range 5–50 % reduction was applied to aluminum sheets produced by accumulative roll bonding to an equivalent strain of 4.8. During tensile testing high resolution maps of the strain...

Pre- and post-irradiation characterization and properties measurements of ZrC coated surrogate TRISO particles

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, Gokul [ORNL; Katoh, Yutai [ORNL; Hunn, John D [ORNL; Snead, Lance Lewis [ORNL

2010-09-01

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in TRISO fuel particles for high temperature gas-cooled reactor fuels. Six sets of ZrC coated surrogate microsphere samples, fabricated by the Japan Atomic Energy Agency using the fluidized bed chemical vapor deposition method, were irradiated in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. These developmental samples available for the irradiation experiment were in conditions of either as-fabricated coated particles or particles that had been heat-treated to simulate the fuel compacting process. Five sets of samples were composed of nominally stoichiometric compositions, with the sixth being richer in carbon (C/Zr = 1.4). The samples were irradiated at 800 and 1250 C with fast neutron fluences of 2 and 6 dpa. Post-irradiation, the samples were retrieved from the irradiation capsules followed by microstructural examination performed at the Oak Ridge National Laboratory's Low Activation Materials Development and Analysis Laboratory. This work was supported by the US Department of Energy Office of Nuclear Energy's Advanced Gas Reactor program as part of International Nuclear Energy Research Initiative collaboration with Japan. This report includes progress from that INERI collaboration, as well as results of some follow-up examination of the irradiated specimens. Post-irradiation examination items included microstructural characterization, and nanoindentation hardness/modulus measurements. The examinations revealed grain size enhancement and softening as the primary effects of both heat-treatment and irradiation in stoichiometric ZrC with a non-layered, homogeneous grain structure, raising serious concerns on the mechanical suitability of these particular developmental coatings as a replacement for SiC in TRISO fuel. Samples with either free carbon or carbon-rich layers dispersed in the ZrC coatings experienced negligible grain size

Material test data of SUS304 welded joints

Energy Technology Data Exchange (ETDEWEB)

Asayama, Tai [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kawakami, Tomohiro [Nuclear Energy System Incorporation, Tokyo (Japan)

1999-10-01

This report summarizes the material test data of SUS304 welded joints. Numbers of the data are as follows: Tensile tests 71 (Post-irradiation: 39, Others: 32), Creep tests 77 (Post-irradiation: 20, Others: 57), Fatigue tests 50 (Post-irradiation: 0), Creep-fatigue tests 14 (Post-irradiation: 0). This report consists of the printouts from 'the structural material data processing system'. (author)

VVER fuel. Results of post irradiation examination

International Nuclear Information System (INIS)

Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

2005-01-01

The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

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

Post-irradiation repairing processes of glucose-6-phosphate dehydrogenase and catalase from Hansenula Polymorpha yeast

International Nuclear Information System (INIS)

Postolache, Carmen; Postolache, Cristian; Dinu, Diana; Dinischiotu, Anca; Sahini, Victor Emanuel

2002-01-01

The post-irradiation repairing mechanisms of two Hansenula Polymorpha yeast enzymes, glucose-6-phosphate dehydrogenase and catalase, were studied. The kinetic parameters of the selected enzymes were investigated over one month since the moment of γ-irradiation with different doses in the presence of oxygen. Dose dependent decrease of initial reaction rates was noticed for both enzymes. Small variation of initial reaction rate was recorded for glucose-6-phosphate dehydrogenase over one month, with a decreasing tendency. No significant electrophoretic changes of molecular forms of this enzyme were observed after irradiation. Continuous strong decrease of catalase activity was evident for the first 20 days after irradiation. Partial recovery process of the catalytic activity was revealed by this study. (authors)

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

The operation of post-irradiation examination facility

International Nuclear Information System (INIS)

Kim, Eun Ka; Min, Duk Ki; Lee, Young Kil

1994-12-01

The operation of post-irradiation examination facility was performed as follow. HVAC and pool water treatment system were continuously operated, and radiation monitoring in PIE facility has been carried out to maintain the facility safely. Inspection of the fuel assembly (F02) transported from Kori Unit 1 was performed in pool, and fuel rods extracted from the fuel assembly (J44) of Kori Unit 2 NPP were examined in hot cell. A part of deteriorated pipe line of drinking water was exchanged for stainless steel pipe to prevent leaking accidents. Halon gas system was also installed in the exhausting blower room for fire fighting. And IAEA inspection camera for safeguard of nuclear materials was fixed at the wall in pool area. Radiation monitoring system were improved to display the area radioactive value at CRT monitor in health physics control room. And automatic check system for battery and emergency diesel generator was developed to measure the voltage and current of them. The performance test of oxide thickness measuring device installed in hot cell for irradiated fuel rod and improvement of the device were performed, and good measuring results using standard sample were obtained. The safeguard inspection of nuclear materials and operation inspection of the facility were carried out through the annual operation inspection, quarterly IAEA inspection and quality assurance auditing. 26 tabs., 43 figs., 14 refs. (Author) .new

SU-E-T-222: Investigation of Pre and Post Irradiation Fading of the TLD100 Thermoluminescence Dosimetry for Photon Beams

Energy Technology Data Exchange (ETDEWEB)

Sina, S [Radiation Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of); Sadeghi, M [Nuclear Engineering department, Shiraz university, Shiraz (Iran, Islamic Republic of); Faghihi, R [Radiation Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Engineering department, Shiraz university, Shiraz (Iran, Islamic Republic of)

2014-06-01

Purpose: The pre-irradiation and post-irradiation fading of the Thermoluminescense dosimeter signals were investigated in this study. Methods: Two groups of TLD chips with pre-determined ECC values were used in this study. The two groups were divided into 6 series, each composing of 5 TLD chips.The first group was used for pre-irradiation fading. 5 TLDs were exposed to a known amount of radiation from Cs-137 source, and were read out the next day. After seven days, the other 5 TLDs were exposed to the same amount of radiation and were read out after a day. The other series of 5 TLDs were also exposed after 7,19,28, 59, and 90 days, and were read out a day after irradiation. The loss in TLD signal were obtained for all the above cases. The second group, was used for postirradiation fading. All the TLDs of this group were exposed to a known amount of radiation from Cs-137 source. The 6 series composed of 5 TLDs were read out after 1,7,19,28,59, and 90 days. The above-mentioned procedures for obtaining pre-irradiation, and post-irradiation fading were performed for three storage temperatures (25°C, 4°C, and −18°C). Results: According to the results obtained in this study, in case of pre-irradiation fading study, the signal losses after 90 days are 12%, 24%, and 17% for 25°C, 4°C, and −18°C respectively. In case of post-irradiation fading study, the sensitivity losses after 90 days are 25%, 216%, and 20% for 25°C, 4°C, and −18°C respectively. Conclusion: The results indicate that the optimized time between exposing and reading out, and also the optimized time between annealing and exposing is 1 day.The reduction of Storage temperature will reduce the post-irradiation fading, While temperature reduction does not have any effect on pre-irradiation fading.

SU-E-T-222: Investigation of Pre and Post Irradiation Fading of the TLD100 Thermoluminescence Dosimetry for Photon Beams

International Nuclear Information System (INIS)

Sina, S; Sadeghi, M; Faghihi, R

2014-01-01

Purpose: The pre-irradiation and post-irradiation fading of the Thermoluminescense dosimeter signals were investigated in this study. Methods: Two groups of TLD chips with pre-determined ECC values were used in this study. The two groups were divided into 6 series, each composing of 5 TLD chips.The first group was used for pre-irradiation fading. 5 TLDs were exposed to a known amount of radiation from Cs-137 source, and were read out the next day. After seven days, the other 5 TLDs were exposed to the same amount of radiation and were read out after a day. The other series of 5 TLDs were also exposed after 7,19,28, 59, and 90 days, and were read out a day after irradiation. The loss in TLD signal were obtained for all the above cases. The second group, was used for postirradiation fading. All the TLDs of this group were exposed to a known amount of radiation from Cs-137 source. The 6 series composed of 5 TLDs were read out after 1,7,19,28,59, and 90 days. The above-mentioned procedures for obtaining pre-irradiation, and post-irradiation fading were performed for three storage temperatures (25°C, 4°C, and −18°C). Results: According to the results obtained in this study, in case of pre-irradiation fading study, the signal losses after 90 days are 12%, 24%, and 17% for 25°C, 4°C, and −18°C respectively. In case of post-irradiation fading study, the sensitivity losses after 90 days are 25%, 216%, and 20% for 25°C, 4°C, and −18°C respectively. Conclusion: The results indicate that the optimized time between exposing and reading out, and also the optimized time between annealing and exposing is 1 day.The reduction of Storage temperature will reduce the post-irradiation fading, While temperature reduction does not have any effect on pre-irradiation fading

Surgical treatments for post-irradiation intestinal injury in uterine cervix cancer patients

International Nuclear Information System (INIS)

Nozaki, Isao; Yokoyama, Nobuji; Takashima, Shigemitsu

1997-01-01

We examined 19 patients with post-irradiation intestinal injury in the uterine cervix cancer for 12 years between 1985 and 1996. We discuss the usefulness and complications of surgery, mainly colostomy. The patients aged from 36 to 80 (average age 61) were treated, and their disease states were 12 cases of rectovaginal fistula, 2 of small intestinal fisfula, 1 of rectum posterior membranous fistula, 3 of proctostenosis, and 14 of proctitis with hemorrhage (including duplication). Surgical methods used were 18 cases of colostomy (2 cases were treated under peritoneum mirror) and 2 of enterocolostomy (including duplication). Eleven out of 19 patients who underwent surgery are alive now. Generally the post-irradiation intestinal injury was intractable, and the method of treatments were limited due to the coexistence of various diseases. The colostomy is safe and less invasive. Therefore patients with uterine cervix cancer having various complications can obtain high quality of life (QOL) such as the improvement of anemia and/or the increase of digestion by the colostomy. (K.H.)

Surgical treatments for post-irradiation intestinal injury in uterine cervix cancer patients

Energy Technology Data Exchange (ETDEWEB)

Nozaki, Isao; Yokoyama, Nobuji; Takashima, Shigemitsu [National Shikoku Cancer Center Hospital, Matsuyama, Ehime (Japan)

1997-06-01

We examined 19 patients with post-irradiation intestinal injury in the uterine cervix cancer for 12 years between 1985 and 1996. We discuss the usefulness and complications of surgery, mainly colostomy. The patients aged from 36 to 80 (average age 61) were treated, and their disease states were 12 cases of rectovaginal fistula, 2 of small intestinal fisfula, 1 of rectum posterior membranous fistula, 3 of proctostenosis, and 14 of proctitis with hemorrhage (including duplication). Surgical methods used were 18 cases of colostomy (2 cases were treated under peritoneum mirror) and 2 of enterocolostomy (including duplication). Eleven out of 19 patients who underwent surgery are alive now. Generally the post-irradiation intestinal injury was intractable, and the method of treatments were limited due to the coexistence of various diseases. The colostomy is safe and less invasive. Therefore patients with uterine cervix cancer having various complications can obtain high quality of life (QOL) such as the improvement of anemia and/or the increase of digestion by the colostomy. (K.H.)

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

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)

DNA replication in ultraviolet light irradiated Chinese hamster cells: the nature of replicon inhibition and post-replication repair

International Nuclear Information System (INIS)

Doniger, J.

1978-01-01

DNA replication in ultraviolet light irradiated Chinese hamster cells was studied using techniques of DNA fiber autoradiography and alkaline sucrose sedimentation. Bidirectionally growing replicons were observed in the autoradiograms independent of the irradiation conditions. After a dose of 5 J/m 2 at 254 nm the rate of fork progression was the same as in unirradiated cells, while the rate of replication was reduced by 50%. After a dose of 10J/m 2 the rate of fork progression was reduced 40%, while the replication rate was only 25% of normal. Therefore, at low doses of ultraviolet light irradiation, the inhibition of DNA replication is due to reduction in the number of functioning replicons, while at higher doses the rate of fork progression is also slowed. Those replicons which no longer function after irradiation are blocked in fork movement rather than replicon initiation. After irradiation, pulse label was first incorporated into short nascent strands, the average size of which was approximately equal to the distance between pyrimidine dimers. Under conditions where post-replication repair occurs these short strands were eventually joined into larger pieces. Finally, the data show that slowing post-replication repair with caffeine does not slow fork movement. The results presented here support the post-replication repair model of 'gapped synthesis' and rule out a major role for 'replicative bypass'. (author)

Modeling of MOS radiation and post irradiation effects

International Nuclear Information System (INIS)

Neamen, D.A.

1984-01-01

The radiation response and long term recovery effects in a n-channel MOSFET due to a pulse of ionizing radiation were modeled assuming that electron tunneling from the semiconductor into the oxide and the buildup of interface states were the postirradiation recovery mechanisms. The modeling used convolution theory and took into account the effects of bias changes during the recovery period and charge yield effects. Changing the bias condition during the post-irradiation recovery period changed the recovery rate. The charge yield effects changed the density of trapped positive charge in the oxide but did not change the recovery characteristics for a given oxide thickness. The modeling results were compared to previous experimental results

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

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

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)

Comparative evaluation of tensile strength of Gutta-percha cones with a herbal disinfectant.

Science.gov (United States)

Mahali, Raghunandhan Raju; Dola, Binoy; Tanikonda, Rambabu; Peddireddi, Suresh

2015-01-01

To evaluate and compare the tensile strength values and influence of taper on the tensile strength of Gutta-percha (GP) cones after disinfection with sodium hypochlorite (SH) and Aloe vera gel (AV). Sixty GP cones of size 110, 2% taper, 60 GP cones F3 ProTaper, and 60 GP of size 30, 6% taper were obtained from sealed packs as three different groups. Experimental groups were disinfected with 5.25% SH and 90% AV gel except the control group. Tensile strengths of GP were measured using the universal testing machine. The mean tensile strength values for Group IA, IIA and IIIA are 11.8 MPa, 8.69 MPa, and 9.24 MPa, respectively. Results were subjected to statistical analysis one-way analysis of variance test and Tukey post-hoc test. 5.25% SH solutions decreased the tensile strength of GP cones whereas with 90% AV gel it was not significantly altered. Ninety percent Aloe vera gel as a disinfectant does not alter the tensile strength of GP cones.

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

Food irradiation - a viable technology for reducing post harvest losses of food

International Nuclear Information System (INIS)

Loaharanu, O.

1985-01-01

Research and development in the past 30 years have clearly demonstrated that food irradiation is a safe, effective and environmentally clean process of food preservation. Twenty-seven countries have approved over 40 irradiated foods or groups of related food items for human consumption, either on an unconditional or a restricted basis. The technology is beginning to play an important role in reducing post-harvest losses of food in facilitating wider distribution of food in the trade. Its wide application in solving microbial spoilage loss of food, insect disinfestation, improving hygenic qualities, slowing down physiological processes of foods is reviewed. Special emphasis is placed on applications of direct relevance to countries in Asia and the Pacific region. (author)

Biochemical effects of heat shock and caffeine on post-irradiation oxic and anoxic damage in barley seeds of low and high water content

International Nuclear Information System (INIS)

Singh, S.P.; Kesavan, P.C.

1991-01-01

Wet heat shock (60 o C, 90s) and caffeine (3.8 x 10 -4 M) afford significant radioprotection against post-irradiation O 2 -dependent damage which develops in seeds of ∼ 3.5% moisture content. The damage was assessed in terms of seedling injury on the eighth day of growth. An increase in seedling injury is clearly seen, associated with a parallel increase in the peroxidase activity. There is a concomitant decrease in the content of total peroxides. Both these post-irradiation treatments potentiate the O 2 -independent component of seedling injury, irrespective of the seed moisture content. Analysis of the peroxidase activity in the seedlings using non-denaturing polyacrylamide gel electrophoresis reveals that two additional bands appear with the post-irradiation oxic damage. Radioprotection against this damage by caffeine, heat shock and O 2 -free post-irradiation hydration is accompanied by the disappearance of these two additional bands. (author)

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

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

Post-Irradiation Behaviour of I131 in TeO2

International Nuclear Information System (INIS)

Jaćimović, Lj.; Stevović, J.; Veljković, S.R.

1965-01-01

The system I 131 in TeO 2 is interesting because little is known about thermal chemical changes in this target. Radioiodine was produced by neutron irradiation of TeO 2 in the reactor. Irradiated TeO 2 was dissolved in diluted NaOH. The analysis of the iodine valency forms was made by ion exchange techniques. The thermal and radiation stability of TeO 2 was studied by using the spectrophotometric method for the determination of tellurium. Post-irradiation annealing of I 131 in TeO 2 was studied in dependence on the time and temperature of the heating. The main tendency of annealing was the reduction of radioiodine. The time dependence of this process indicates a fast change at high temperatures. The curves are more complex at lower temperatures. The annealing may appear complex because of the variety of thermal reactions of iodine intermediary. It may react with products of the following processes: tellurium recoil and corresponding hot zone, beta transition of Te 131 and TeO 2 itself. The kinetics of these changes was considered and an estimation of the processes during annealing was made. The influence of the neutron flux on the kinetics of annealing was also studied. (author) [fr

Post-irradiation degradation of DNA in electron and neutron-irradiated E. coli B/r; the effect of the radiation sensitizer metronidazole

Energy Technology Data Exchange (ETDEWEB)

Cramp, W A; George, A M; Howlett, J [Hammersmith Hospital, London (UK). M.R.C. Cyclotron Unit

1976-04-01

Suspensions of E.coli B/r were irradiated under aerobic and anoxic conditions with electrons (7 to 8 MeV, 2 and 20 krad/min, MRC linear accelerator), or with neutrons (average energy 7.5 MeV, 2 krad/min, MRC cyclotron) in an investigation of the effects of the radiosensitizer, metronidazole (Flagyl, 5 or 10 mM) on survival and DNA degradation. These results are compared with those for another electron affinic radiosensitizer, indane trione. Survival studies yielded enhancement ratios, for anoxic irradiation only, of 1.7 (5mM) and 1.9 (10mM) for electrons, and 1.2 (5mM and 10mM) for neutrons. Unlike indane trione, metronidazole had no pronounced inhibitory effect on post-irradiation DNA degradation, either when incubated with the bacteria before irradiation or when present during irradiation. When present under anoxic conditions of irradiation with electrons, some enhancement of degradation was observed. DNA degradation was reduced at higher doses, with a pronounced maxiumum effect, for neutrons as well as for electrons. Metronidazole allowed this degradation to continue and showed some sensitizing action, but did not prevent the decrease in total degradation at high doses. It is therefore difficult to correlate DNA degradation with cell-depth.

New radiation mitigators to reduce bone marrow death of mice by post-irradiation administration

International Nuclear Information System (INIS)

Anzai, Kazunori

2009-01-01

We have found recently that heat-treated mineral yeast preparations and water-soluble analogs of vitamin E are potent radiation mitigator to reduce bone marrow death of mice by post-irradiation administration. When administered immediately after whole-body X-irradiation (7.5 Gy), both Zn-yeast and γ-tocopherol dimethylglycine ester (TDMG) significantly increased the viability of mice from 0% (control) to more than 90% (treated). Zn-yeast did not inhibit the tumor-regulation by γ-rays but even sensitize the radiation effect in mice xenografts of HeLa cells. (author)

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

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, E. G.; Jeon, Y. B.; Ku, D. S.

1996-12-01

In 1996, the post-irradiation examination(PIE) of nuclear fuels was performed as follows. It has been searched for the caution of defection of defected fuel rods of Youngkwang-4 reactor through NDT and metallographic examination that had been required by KEPCO. And in-pool inspection of Kori-1 spent fuel assembly(FO2) was carried out. HVAC system and pool water treatment system have been operated to maintain the facility safely, and electric power supply system was checked and maintained for the normal and steady supply electric power to the facility. Image processing software was developed for measurement of defection of spent fuel rods. Besides, a radiation shielding glove box was fabricated and a hot cell compressor for volume reduction of radioactive materials was fabricated and installed in hot cell. Safeguards of nuclear materials were implemented in strict accordance with the relevant Korean rules and regulations as well as the international non-proliferation regime. Also the IAEA inspection was carried out on the quarterly basis. (author). 31 tabs., 71 figs., 4 refs.

Biochemical effects of heat shock and caffeine on post-irradiation oxic and anoxic damage in barley seeds of low and high water content

Energy Technology Data Exchange (ETDEWEB)

Singh, S.P.; Kesavan, P.C. (Jawaharlal Nehru Univ., New Delhi (India). School of Life Sciences)

1991-05-01

Wet heat shock (60{sup o}C, 90s) and caffeine (3.8 x 10{sup -4}M) afford significant radioprotection against post-irradiation O{sub 2}-dependent damage which develops in seeds of {similar to} 3.5% moisture content. The damage was assessed in terms of seedling injury on the eighth day of growth. An increase in seedling injury is clearly seen, associated with a parallel increase in the peroxidase activity. There is a concomitant decrease in the content of total peroxides. Both these post-irradiation treatments potentiate the O{sub 2}-independent component of seedling injury, irrespective of the seed moisture content. Analysis of the peroxidase activity in the seedlings using non-denaturing polyacrylamide gel electrophoresis reveals that two additional bands appear with the post-irradiation oxic damage. Radioprotection against this damage by caffeine, heat shock and O{sub 2}-free post-irradiation hydration is accompanied by the disappearance of these two additional bands. (author).

The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

International Nuclear Information System (INIS)

Roettger, R.; Bueger, J. de; Schoots, T.

1977-01-01

A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

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.))

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

International Nuclear Information System (INIS)

Byun, T.S.; Farrell, K.; Lee, E.H.; Mansur, L.K.; Maloy, S.A.; James, M.R.; Johnson, W.R.

2002-01-01

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 0.54-2.53 dpa at 30-100 deg. C. Tensile testing was performed at room temperature (20 deg. C) and 164 deg. C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 deg. C decreased the strength by 13-18% and the ductility by 8-36%. 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. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability

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)

Post-cracking tensile behaviour of steel-fibre-reinforced roller-compacted-concrete for FE modelling and design purposes

International Nuclear Information System (INIS)

Jafarifar, N.; Pilakoutas, K.; Angelakopoulos, H.; Bennett, T.

2017-01-01

Fracture of steel-fibre-reinforced-concrete occurs mostly in the form of a smeared crack band undergoing progressive microcracking. For FE modelling and design purposes, this crack band could be characterised by a stress-strain (σ-ε) relationship. For industrially-produced steel fibres, existing methodologies such as RILEM TC 162-TDF (2003) propose empirical equations to predict a trilinear σ-ε relationship directly from bending test results. This paper evaluates the accuracy of these methodologies and their applicability for roller-compacted-concrete and concrete incorporating steel fibres recycled from post-consumer tyres. It is shown that the energy absorption capacity is generally overestimated by these methodologies, sometimes up to 60%, for both conventional and roller-compacted concrete. Tensile behaviour of fibre-reinforced-concrete is estimated in this paper by inverse analysis of bending test results, examining a variety of concrete mixes and steel fibres. A multilinear relationship is proposed which largely eliminates the overestimation problem and can lead to safer designs. [es

Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

International Nuclear Information System (INIS)

Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

2010-01-01

An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

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

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

Minimising contralateral breast dose in post-mastectomy intensity-modulated radiotherapy by incorporating conformal electron irradiation

NARCIS (Netherlands)

van der Laan, Hans Paul; Korevaar, Erik W; Dolsma, Willemtje; Maduro, John H; Langendijk, Johannes A

PURPOSE: To assess the potential benefit of incorporating conformal electron irradiation in intensity-modulated radiotherapy (IMRT) for loco-regional post-mastectomy RT. PATIENTS AND METHODS: Ten consecutive patients that underwent left-sided mastectomy were selected for this comparative planning

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.)

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

Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

Energy Technology Data Exchange (ETDEWEB)

Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

2016-01-15

Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

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.)

Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

International Nuclear Information System (INIS)

Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

1996-01-01

Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

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)

Post-factum detection of food irradiation

International Nuclear Information System (INIS)

Meier, W.

1991-01-01

Irradiation of food containing bones or shells can be detected with a high degree of certainty by means of ESR and by analysis of the volatile hydrocarbons or of the o-tyrosine. The last two methods are used for identification of irradiated pure meat samples. Detection of irradiation in spices and dried vegetables is possible with the thermoluminescence method and ESR, if non-irradiated control samples are available. These methods are being tested in the period 1990/1991 by an EC Commission-sponsored interlaboratory study of spices and food containing bones or shells, whereas the two chemical methods need further optimisation by work done in smaller working groups. (orig.) [de

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)

Development of a miniaturized bulge test (small punch test) for post-irradiation mechanical property evaluation

International Nuclear Information System (INIS)

Eto, Motokuni; Suzuki, Masahide; Nishiyama, Yutaka; Fukaya, Kiyoshi; Jitsukawa, Shiro; Misawa, Toshihei

1993-01-01

To examine the effectiveness of the small punch test for evaluating strength and toughness of irradiated ferritic steels, detailed procedures are described aiming at standardization of the test. The statistical approach to analysis of the SP energy as a function of temperature for evaluation of DBTT was also reviewed. The method was then applied to neutron-irradiated ferritic steels, which included F-82, F-82H, HT-9, and 2 1/4 Cr-1Mo steel. Fluence and irradiation temperatures ranged from 2 to 12 x 10 23 n/m 2 (E ≥ 1 MeV) and from 573 to 673 K, respectively. Comparison of parameters obtained from the small punch test with the properties measured by the conventional method indicated that: (a) the 0.2% offset stress and the ultimate tensile strength at room temperature can be correlated well with the parameters, P y /(t 0 ) 2 and P max /(t 0 ) 2 , respectively. Here, P y and P max are the loads corresponding to the yield and the maximum, and t 0 is the initial thickness of a specimen; (b) fracture toughness, J IC , can be evaluated using equivalent fracture strain, anti ε qf , and the previously established relationship between these values; and (c) DBTT measured by a Charpy test can be predicted from the results of temperature dependence of SP energy determined from the area under the load-deflection curve using a statistical analysis based on a Weibull distribution

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

2017-08-15

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

Directory of Open Access Journals (Sweden)

Jong Man Park

2017-08-01

Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

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.

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.

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

DAMAGE IN MOLYBDENUM ASSOCIATED WITH NEUTRON IRRADIATION AND SUBSEQUENT POST-IRRADIATION ANNEALING

Energy Technology Data Exchange (ETDEWEB)

Mastel, B.

1963-07-23

Molybdemum containing carbon was studied in an attempt to establish the combined effect of impurity content and neutron irradiation on the properties and structure of specific metals. Molybdenum foils were punched into discs and heat treated in vacuum. They were then slow-cooled and irradiated. After irradiation and subsequent decay of radioactivity to a low level the foils were subjected to x-ray diffraction measurements. Cold-worked foils with less than 10 ppm carbon showed no change in microstructure due to irradiation. Molybdenum foils that were annealed prior to irradiation showed spot defects. In foils containing up to 500 ppm carbon, it was concluded that the small loops present after irradiation are due to the clustering of point defects at interstitial carbon atoms, followed by collapse to form a dislocation loop. The amount of lattice expansion after irradiation was strongly dependent on impurity content. Neutron irradiation was found to reduce the number of active slip systems. (M.C.G.)

Free radical generation in post-irradiation period: an evidence from the conversion of xanthine dehydrogenase into xanthine oxidase

International Nuclear Information System (INIS)

Kale, R.K.

2003-01-01

Xanthine oxidoreductase (XOR) system which consists of xanthine dehydrogenase (XDH) and xathine oxidase (XO), is one of the major sources of free radicals in biological systems. XOR system is pre-dominantly present as XDH in the normal tissue and converts into free radical generating XO-Form in the damaged tissue. Therefore, XO-Form of XOR system, is expected to be mainly found in the radiolytically damaged tissue. In such an event, XO may catalyze the generation of free radicals and potentiate the radiation effects in post-irradiation period. Recent findings on the effect of ionizing radiation on XOR system in the liver of mice, peroxidative damage and lactate dehydrogenase support this possibility. From these results it has been hypothized that free radical generating systems could be activated in the radiolytically damaged cell and in turn contribute to the cause and complications of late effects and their persistence in post-irradiation period. This aspect may have great significance in understanding the radiation - induced damages. It may also have serious implication in various fields like radiation therapy, health physics, carcinogenesis, space travelling radiation exposures and post nuclear accident care. Further, it is suggested that efforts need to be made to search more system(s) which could be activated particularly at lower doses of radiation and generate free radicals in post-irradiation period

Post-irradiation vasculopathy of intracranial major arteries in children; Report of two cases

Energy Technology Data Exchange (ETDEWEB)

Nishizawa, Shigeru; Ryu, Hiroshi; Yokoyama, Tetsuo; Ninchoji, Toshiaki; Shimoyama, Ichiro; Yamamoto, Seiji; Uemura, Kenichi [Hamamatsu Univ. School of Medicine, Shizuoka (Japan)

1991-06-01

We report two rare cases of post-irradiation vasculopathy of intracranial major arteries in children. A 13-year-old girl suffered from transient right hemiparesis 1 year after irradiation for suprasellar germinoma. Left carotid angiograms revealed marked stenoses of the intracranial internal carotid, middle cerebral, and anterior cerebral arteries, which were previously normal, and moyamoya vessels. A 2.5-year-old girl underwent internal irradiation with {sup 198}Au colloid for cystic craniopharyngioma. At the age of 10 years, she suddenly became unconscious after vomiting. Computed tomographic scans showed a right frontal intracerebral hematoma. Right carotid angiograms disclosed complete obstruction of the intracranial internal carotid, middle cerebral, and anterior cerebral arteries and moyamoya vessels, previously not present. The danger of radiation therapy causing occlusive vasculopathy in small and major cerebral arteries in children is emphasized. To prevent permanent ischemic neurological deficits, vasculopathy should be treated either medically or surgically as early as possible. (author).

Post-X-irradiation effects on petunia pollen germinating in vitro and in vivo

International Nuclear Information System (INIS)

Gilissen, L.J.W.

1978-01-01

The germination of Petunia hybrida L. pollen grains in germination medium, containing 10% sucrose and 0.01 % H 3 BO 3 , was linearly related to relative humidity (RH): being minimal at 0 % RH and maximal at 100 % RH. The low germination at 0 % RH was completely restored after transfer to 100 % RH. Germination in medium decreased with increasing X-ray exposures between O and 400 kR. This decrease was caused by pollen rupture. No in vitro germination occurred at exposures of 400 kR and more. The radiosensitivity of pollen in vitro was minimal at 80 % RH. Transfer of pollen to the stigma post-X-irradiation resulted in resistance to much higher exposures of irradiation (<750 kR). The differences in radiosensitivity of the pollen germinated in vitro and in vivo are due possibly to the differences in composition of the germination medium and the stigmatic exudate. Pollen tube growth of irradiated pollen after compatible or incompatible pollination at first showed retarded then normal tube growth. A conclusion is that X-irradiation of pollen cannot influence the characteristics of pollen tube growth after compatible or incompatible pollination. (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.

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)

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.

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.

State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

International Nuclear Information System (INIS)

Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

2015-01-01

The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

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

Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

1997-01-01

Three prototype fuel rods containing Al-64 wt% U 3 Si 2 (3.15 gU/cm 3 ) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U 3 Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U 3 Si 2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U 3 Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

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.)

In vitro tensile strength of luting cements on metallic substrate.

Science.gov (United States)

Orsi, Iara A; Varoli, Fernando K; Pieroni, Carlos H P; Ferreira, Marly C C G; Borie, Eduardo

2014-01-01

The aim of this study was to determine the tensile strength of crowns cemented on metallic substrate with four different types of luting agents. Twenty human maxillary molars with similar diameters were selected and prepared to receive metallic core castings (Cu-Al). After cementation and preparation the cores were measured and the area of crown's portion was calculated. The teeth were divided into four groups based on the luting agent used to cement the crowns: zinc phosphate cement; glass ionomer cement; resin cement Rely X; and resin cement Panavia F. The teeth with the crowns cemented were subjected to thermocycling and later to the tensile strength test using universal testing machine with a load cell of 200 kgf and a crosshead speed of 0.5 mm/min. The load required to dislodge the crowns was recorded and converted to MPa/mm(2). Data were subjected to Kruskal-Wallis analysis with a significance level of 1%. Panavia F showed significantly higher retention in core casts (3.067 MPa/mm(2)), when compared with the other cements. Rely X showed a mean retention value of 1.877 MPa/mm(2) and the zinc phosphate cement with 1.155 MPa/mm(2). Glass ionomer cement (0.884 MPa/mm(2)) exhibited the lowest tensile strength value. Crowns cemented with Panavia F on cast metallic posts and cores presented higher tensile strength. The glass ionomer cement showed the lowest tensile strength among all the cements studied.

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

A Study on the Pre-and Post-irradiation Effect of Blood Vessels in the Experimentally Induced Tongue Cancer

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Tae; Park, Tae Won [Dept. of Oral Radiology, College of Dentistry, Seoul National University, Seoul (Korea, Republic of)

1990-02-15

The author observed the changes of vasculature of pre-and post-irradiation on DMBA induced rat tongue cancer. The study was performed by using vascular corrosion resin casting, and scanning electron microscopy. The results were as follows. 1. The capillaries runned parallely and formed bundles and, sometimes, plexus. The endothelial cells were arranged regularly and small pores were observed. 2. In irradiated normal tongue the capillaries were curved slightly and formed plexus on initial day of post-irradiation. On third day the capillaries and capillary pores were dilated and the endothelial cell arrangement was irregular. The effects of irradiation were gradually increased from initial to the 3rd day, though it was decreased after 7th day. 3. The vasculature of DMBA induced tongue cancer group were very irregular, and large avascular lesions were formed according to the cancer necrosis or tumor cell nest and the vasculature was narrowed and paralleled around the avascular lesion by compression of cancer cell nest. The vascular wall was roughened and dilated, forming club shaped or varix. 4. The vessels were curved and formed reticular network in irradiated DMBA induced tongue carinoma group. The free end of newly formed capillaries had regular width, and also irregular club shaped or aneurysmal dilation were observed. The vascular structures were destroyed and vessels were fused in tumor necrosis lesion. The radiation effects were marked on the first and third day of irradiation and the effects were decreased after seventh day and showed capillary regeneration.

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)

Patterning of gold nano-octahedra using electron irradiation combined with thermal treatment and post-cleaning process

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Nam; Kum, Jong Min [Korea Advanced Institute of Science and Technology (KAIST), Department of Nuclear and Quantum Engineering (Korea, Republic of); Lee, Hyeok Moo [Korea Atomic Energy Research Institute (KAERI), Research Division for Industry and Environment (Korea, Republic of); Cho, Sung Oh, E-mail: socho@kaist.ac.kr [Korea Advanced Institute of Science and Technology (KAIST), Department of Nuclear and Quantum Engineering (Korea, Republic of)

2012-03-15

A novel approach to pattern nanocrystalline gold (Au) octahedra is presented based on electron irradiation combined with thermal treatment and post-cleaning process using HAuCl{sub 4}-loaded poly(styrene-b-2-vinyl pyridine) (PS-b-P2VP) block copolymer (BCP) as a precursor material. The BCP tends to cross-link under electron irradiation, and thus a patterned film can be prepared by selectively irradiating an electron beam onto a precursor film using a shadow mask. A post-thermal treatment leads to the formation of crystalline Au nano-octahedra inside the patterned film with a help of the BCP acting as a capping agent. Subsequently, the BCP can be removed by O{sub 2} plasma etching combined with oxidative degradation, with the Au nanoparticles remaining. As a result, a patterned film consisting of high-purity nanocrystalline Au octahedra is fabricated. The sizes of the Au octahedral nanoparticles can be readily controlled from 49 to 101 nm by changing the thickness of the precursor film. The patterned Au nano-octahedra films exhibit excellent surface-enhanced Raman scattering behavior with the maximum enhancement factor of {approx}10{sup 6}.

Alkali-labile sites and post-irradiation effects in single-stranded DNA induced by H radicals

International Nuclear Information System (INIS)

Lafleur, M.V.M.; Heuvel, N. van; Woldhuis, J.; Loman, H.

1978-01-01

Single-stranded phiX174 DNA in aqueous solutions has been irradiated in the absence of oxygen, under conditions in which H radicals react with the DNA. It was shown that H radical reactions result in breaks, which contribute approximately 10 per cent inactivation. Further, two types of alkali-labile sites were formed. One was lethal and gave rise to single-strand breaks by alkali and was most probably identical with post-irradiation heat damage and contributed about 33 per cent to the inactivation mentioned above. The other consisted of non-lethal damage, partly dihydropyrimidine derivatives, and was converted to lethal damage by alkali. This followed from experiments in which the DNA was treated with osmium-tetroxide, which oxidized thymine to 5,6-dihydroxydihydrothymine. Treatment with alkali of this DNA gave the same temperature dependence as found for the non-lethal alkali-labile sites in irradiated DNA. A similar temperature dependence was found for dihydrothymine and irradiated pyrimidines with alkali. (author)

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

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)

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)

Program description for the qualification of CNEA - Argentina as a supplier of LEU silicide fuel and post-irradiation examinations plan for the first prototype irradiated in Argentina

International Nuclear Information System (INIS)

Rugirello, Gabriel; Adelfang, Pablo; Denis, Alicia; Zawerucha, Andres; Marco, Agustin di; Guillaume, Eduardo; Sbaffoni, Monica; Lacoste, Pablo

1998-01-01

In this report we present a description of the ongoing and future stages of the program for the qualification of CNEA, Argentina, as a supplier of low enriched uranium silicide fuel elements for research reactor. Particularly we will focus on the characteristics of the future irradiation experiment on a new detachable prototype, the post-irradiation examinations (PIE) plan for the already irradiated prototype PO4 and an overview of the recently implemented PIE facilities and equipment. The program is divided in several steps, some of which have been already completed. It concludes: development of the uranium silicide fissile material, irradiation and PIE of several full-scale prototypes. Important investments have been already carried out in the facilities for the FE production and PIE. (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)

Polyamines and post-irradiation cell proliferation

International Nuclear Information System (INIS)

Rosiek, O.; Wronowski, T.; Lerozak, K.; Kopec, M.

1978-01-01

The results of three sets of experiments will be presented. Firstly polyamines and DNA content was determined in bone marrow, mesenteric lymph nodes, spleen, liver and kidney of rabbits at the 1, 5, 10 and 20th day after exposure to 600 R of X-irradiation. Polyamine concentration in bone marrow, spleen and lymph nodes was found to be markedly increased during the period of postirradiation recovery. Secondly, effect of 10 -5 M methyl glyoxalbis, guanylhydrazone (MGBG), an inhibitor of spermidine and spermine synthesis, on multiplication of X-irradiated cultures of murine lymphoblaste L5178Y-S was assessed. MGBG-induced inhibition of cell proliferation could be prevented by concurrent administration of 10 -4 M spermidine. Thirdly the influence of putrescine on bone marrow cellularity and 3 H-thymidine incorporation into bone marrow cells was investigated in X-irradiated mice. The results obtained indicate close relation of polyamines to cell proliferation processes after irradiation. (orig./AJ) [de

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

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

Directory of Open Access Journals (Sweden)

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.

Investigation of the effect of some irradiation parameters on the response of various types of dosimeters to electron irradiation

International Nuclear Information System (INIS)

Farah, K.; Kuntz, F.; Kadri, O.; Ghedira, L.

2004-01-01

Several undyed and dyed polymer films are commercially available for dosimetry in intense radiation fields, especially for radiation processing of food and sterilisation of medical devices. The effects of temperature during irradiation and post-irradiation stability, on the response of these dosimeters are of importance to operators of irradiation facilities. The present study investigates the effects of temperature during irradiation by 2.2 MeV electrons beam accelerator and post irradiation storage on the response of several types of dosimeter films. All dosimeters showed a significant effect of temperature during irradiation and post-irradiation storage

Investigation of the effect of some irradiation parameters on the response of various types of dosimeters to electron irradiation

Energy Technology Data Exchange (ETDEWEB)

Farah, K. E-mail: k.farah@cnstn.rnrt.tn; Kuntz, F.; Kadri, O.; Ghedira, L

2004-10-01

Several undyed and dyed polymer films are commercially available for dosimetry in intense radiation fields, especially for radiation processing of food and sterilisation of medical devices. The effects of temperature during irradiation and post-irradiation stability, on the response of these dosimeters are of importance to operators of irradiation facilities. The present study investigates the effects of temperature during irradiation by 2.2 MeV electrons beam accelerator and post irradiation storage on the response of several types of dosimeter films. All dosimeters showed a significant effect of temperature during irradiation and post-irradiation storage.

The importance of using the irradiation technology in the post-harvest Preservation of onions and garlic

International Nuclear Information System (INIS)

Iglesias Enriquez, Isora

1999-01-01

In Cuba post-harvest preservation of onions and garlic for different uses have been performed by irradiation bulbs with a minimal dose range of 80 to 90 Gy of Gamma radiation (Co60 ) at commercial level in the Food Irradiation Plant (PIA) Producto 1 which in 1986 held a nominal activity of 110 000 ci. Results showed that the irradiated products could be preserved up to 8 and 11 months, respectively, resulting un total losses lower than 30 %. Products were stored in a warehouse with forced air distribution system of 22 0C to 32 0C and 70 to 100 % RH, resulting in 30 air changes /hour. An important economic benefit was obtained from this method as compared to other traditional storage methods using controlled temperature chambers ( 1 0C to 3 0C ) to preserve un-irradiated onions an garlic's. It is concluded that the irradiated products could be stored at atmospheric temperature and forced air distribution system resulting in lower losses and energy savings and non-imported product, which to reached more of the 5 dollars millions

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)

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.

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

Directory of Open Access Journals (Sweden)

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

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

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.

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

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

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

Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

International Nuclear Information System (INIS)

Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

2002-01-01

As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

International Nuclear Information System (INIS)

Karam, M.; Dimayuga, F.C.; Montin, J.

2010-01-01

AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O 2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt.% Th and 1.53 wt.% Pu in (Th, Pu)O 2 . The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O 2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O 2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O 2 fuel performance characteristics were superior to UO 2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

Radioprotection by caffeine pre-treatment and post-treatment in the bone marrow chromosomes of mice given whole-body γ-irradiation

International Nuclear Information System (INIS)

Farooqi, Z.; Kesavan, P.C.

1992-01-01

The effect of caffeine given as pre- and post-treatment in mice exposed to whole-body γ-irradiation (1.5 Gy 60 Co γ-rays) was studied. The pre-treatment was either acute or chronic. The acute dose (5 mg/kg and 15 mg/kg body weight) was in the form of an injection given intraperitoneally, 30 min before irradiation. The chronic administration was in the form of caffeine solution (4.208x10 -3 M and 7.72x10 -4 M) contained in drinking water for 5 weeks prior to radiation exposure. The acute pre-treatment with caffeine reduced the radiation-induced frequency of chromosomal aberrations discernibly, whereas chronic pre-treatment afforded a much more significant degree of radioprotection. The caffeine post-treatment (5 mg/kg and 15 mg/kg body weight) was given in the form of an intraperitoneal injection to the mice immediately following whole-body γ-irradiation. It is noted that both post-treatment concentrations of caffeine also significantly reduced the frequency of chromosomal aberrations induced by γ-rays. These data are briefly discussed in terms of possible mechanistic considerations. (author). 33 refs.; 3 tabs

Radioprotection by caffeine pre-treatment and post-treatment in the bone marrow chromosomes of mice given whole-body [gamma]-irradiation

Energy Technology Data Exchange (ETDEWEB)

Farooqi, Z.; Kesavan, P.C. (Jawaharlal Nehru Univ., New Delhi (India). School of Life Sciences)

1992-10-01

The effect of caffeine given as pre- and post-treatment in mice exposed to whole-body [gamma]-irradiation (1.5 Gy [sup 60]Co [gamma]-rays) was studied. The pre-treatment was either acute or chronic. The acute dose (5 mg/kg and 15 mg/kg body weight) was in the form of an injection given intraperitoneally, 30 min before irradiation. The chronic administration was in the form of caffeine solution (4.208x10[sup -3] M and 7.72x10[sup -4] M) contained in drinking water for 5 weeks prior to radiation exposure. The acute pre-treatment with caffeine reduced the radiation-induced frequency of chromosomal aberrations discernibly, whereas chronic pre-treatment afforded a much more significant degree of radioprotection. The caffeine post-treatment (5 mg/kg and 15 mg/kg body weight) was given in the form of an intraperitoneal injection to the mice immediately following whole-body [gamma]-irradiation. It is noted that both post-treatment concentrations of caffeine also significantly reduced the frequency of chromosomal aberrations induced by [gamma]-rays. These data are briefly discussed in terms of possible mechanistic considerations. (author). 33 refs.; 3 tabs.

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.

Molecular structure effects on the post irradiation diffusion in polymer gel dosimeters

Energy Technology Data Exchange (ETDEWEB)

Mattea, F.; Romero, M.; Strumia, M. [Instituto Multidisciplinario de Biologia Vegetal / CONICET, Universidad Nacional de Cordoba, Departamento de Quimica Organica, Ciudad Universitaria, 5000 Cordoba (Argentina); Vedelago, J. [Laboratorio de Investigaciones e Instrumentacion en Fisica Aplicada a la Medicina e Imagenes por Rayos X, Laboratorio 448 FaMAF - UNC, Ciudad Universitaria, 5000 Cordoba (Argentina); Quiroga, A. [Centro de Investigacion y Estudios de Matematica / CONICET, Oficina 318 FaMAF - UNC, Ciudad Universitaria, 5000 Cordoba (Argentina); Valente, M., E-mail: fmattea@gmail.com [Instituto de Fisica E. Gaviola / CONICET, LIIFAMIRx, Oficina 102 FaMAF - UNC, 5000 Cordoba (Argentina)

2014-08-15

Polymer gel dosimeters have specific advantages for recording 3D radiation dose distribution representing a key factor for most of the therapeutic and diagnostic radiation techniques. Radiation-induced polymerization and crosslinking reactions that take place in the dosimeter have been studied for different monomers like acrylamide and N,N-methylene-bis acrylamide (Bis) and most recently for less toxic monomers like N-isopropylacrylamide and Bis. In this work a novel system based on itaconic acid and Bis is proposed, the radical polymerization or gel formation of these monomers has been already studied for the formation of an hydrogel for non dosimetric applications and their reactivity are comparable with the already mentioned systems. Although the 3D structure is maintained after the dosimeter has been irradiated, it is not possible to eliminate the diffusion of the reacted and monomer species in regions of dose gradients within the gel after irradiation. As a consequence the dose information of the dosimeters loose quality over time. The mobility within the gelatin structure of the already mentioned species is related to their chemical structure, and nature. In this work the effect of changes in the chemical structure of the monomers over the dosimetric sensitivity and over the post-irradiation diffusion of species is studied. One of the acrylic acid groups of the itaconic acid molecule is modified to obtain molecules with similar reactivity but different molecular sizes. Dosimetric systems with these modified species, Bis, an antioxidant to avoid oxygen polymerization inhibition, water and gelatin are irradiated in an X-ray tomography at different doses, and the resulting dosimeters are characterized by Raman spectroscopy and optical absorbance to study their feasibility and capabilities as dosimetric systems, and by optical-CT to analyze the diffusion degree after being irradiated. (Author)

Molecular structure effects on the post irradiation diffusion in polymer gel dosimeters

International Nuclear Information System (INIS)

Mattea, F.; Romero, M.; Strumia, M.; Vedelago, J.; Quiroga, A.; Valente, M.

2014-08-01

Polymer gel dosimeters have specific advantages for recording 3D radiation dose distribution representing a key factor for most of the therapeutic and diagnostic radiation techniques. Radiation-induced polymerization and crosslinking reactions that take place in the dosimeter have been studied for different monomers like acrylamide and N,N-methylene-bis acrylamide (Bis) and most recently for less toxic monomers like N-isopropylacrylamide and Bis. In this work a novel system based on itaconic acid and Bis is proposed, the radical polymerization or gel formation of these monomers has been already studied for the formation of an hydrogel for non dosimetric applications and their reactivity are comparable with the already mentioned systems. Although the 3D structure is maintained after the dosimeter has been irradiated, it is not possible to eliminate the diffusion of the reacted and monomer species in regions of dose gradients within the gel after irradiation. As a consequence the dose information of the dosimeters loose quality over time. The mobility within the gelatin structure of the already mentioned species is related to their chemical structure, and nature. In this work the effect of changes in the chemical structure of the monomers over the dosimetric sensitivity and over the post-irradiation diffusion of species is studied. One of the acrylic acid groups of the itaconic acid molecule is modified to obtain molecules with similar reactivity but different molecular sizes. Dosimetric systems with these modified species, Bis, an antioxidant to avoid oxygen polymerization inhibition, water and gelatin are irradiated in an X-ray tomography at different doses, and the resulting dosimeters are characterized by Raman spectroscopy and optical absorbance to study their feasibility and capabilities as dosimetric systems, and by optical-CT to analyze the diffusion degree after being irradiated. (Author)

IAEA Post Irradiation Examination Facilities Database

International Nuclear Information System (INIS)

Jenssen, Haakon; Blanc, J.Y.; Dobuisson, P.; Manzel, R.; Egorov, A.A.; Golovanov, V.; Souslov, D.

2005-01-01

The number of hot cells in the world in which post irradiation examination (PIE) can be performed has diminished during the last few decades. This creates problems for countries that have nuclear power plants and require PIE for surveillance, safety and fuel development. With this in mind, the IAEA initiated the issue of a catalogue within the framework of a coordinated research program (CRP), started in 1992 and completed in 1995, under the title of ''Examination and Documentation Methodology for Water Reactor Fuel (ED-WARF-II)''. Within this program, a group of technical consultants prepared a questionnaire to be completed by relevant laboratories. From these questionnaires a catalogue was assembled. The catalogue lists the laboratories and PIE possibilities worldwide in order to make it more convenient to arrange and perform contractual PIE within hot cells on water reactor fuels and core components, e.g. structural and absorber materials. This catalogue was published as working material in the Agency in 1996. During 2002 and 2003, the catalogue was converted to a database and updated through questionnaires to the laboratories in the Member States of the Agency. This activity was recommended by the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) at its plenary meeting in April 2001. The database consists of five main areas about PIE facilities: acceptance criteria for irradiated components; cell characteristics; PIE techniques; refabrication/instrumentation capabilities; and storage and conditioning capabilities. The content of the database represents the status of the listed laboratories as of 2003. With the database utilizing a uniform format for all laboratories and details of technique, it is hoped that the IAEA Member States will be able to use this catalogue to select laboratories most relevant to their particular needs. The database can also be used to compare the PIE capabilities worldwide with current and future

Isothermal and isochronal annealing methodology to study post-irradiation temperature activated phenomena

International Nuclear Information System (INIS)

Chabrerie, C.; Autran, J.L.; Paillet, P.; Flament, O.; Leray, J.L.; Boudenot, J.C.

1997-01-01

In this work, the evolution of the oxide trapped charge has been modeled, to predict post-irradiation behavior for arbitrary anneal conditions (i.e., arbitrary temperature-time profiles). Using experimental data obtained from a single isochronal anneal, the method consists of calculating the evolution of the energy distribution of the oxide trapped charge, in the framework of a thermally activated charge detrapping model. This methodology is illustrated in this paper by the prediction of experimental isothermal data from isochronal measurements. The implications of these results to hardness assurance test methods are discussed

A data recording and processing system for examination of irradiated fuel rods

International Nuclear Information System (INIS)

Olshausen, K.D.; Christiansen, J.F.; Larsen, J.K.; Stevens, T.

1978-03-01

Over the past eight years a system has been developed at Institutt for atomenergi, Kjeller, Norway, which helps to collect and analyze data obtained during certain post-irradiation examination techniques employed at IFA's Hot Laboratory. At present computer codes are in use for treating results from: profilometry, gamma scanning, eddy current testing and tensile testing. The system is easily adaptable to any othermeasuring data with the structure: -axial position, time interval etc.; signal A (DC-voltage, counts); signal B (DC-voltage, counts). Anextention of the system for transverse gamma scanning and gap measurement (mechanical compression method) is planned for the future. The various programmes in the system are described and examples of the outputs are given in a series of appendices. (JIW)